Fusion-Based Neutron Generator Production of Tc-99m and Tc-101: A Prospective Avenue to Technetium Theranostics

Presented are the results of 99mTc and 101Tc production via neutron irradiation of natural isotopic molybdenum (Mo) with epithermal/resonance neutrons. Neutrons were produced using a deuterium-deuterium (D-D) neutron generator with an output of 2 × 1010 n/s. The separation of Tc from an irradiated source of bulk, low-specific activity (LSA) Mo on activated carbon (AC) was demonstrated. The yields of 99mTc and 101Tc, together with their potential use in medical single-photon emission computed tomography (SPECT) procedures, have been evaluated from the perspective of commercial production, with a patient dose consisting of 740 MBq (20 mCi) of 99mTc. The number of neutron generators to meet the annual 40,000,000 world-wide procedures is estimated for each imaging modality: 99mTc versus 101Tc, D-D versus deuterium-tritium (D-T) neutron generator system outputs, and whether or not natural molybdenum or enriched targets are used for production. The financial implications for neutron generator production of these isotopes is also presented. The use of 101Tc as a diagnostic, therapeutic, and/or theranostic isotope for use in medical applications is proposed and compared to known commercial nuclear diagnostic and therapeutic isotopes.


Introduction
99m Tc (t 1/2 = 6.007 h; IT 99.9%) is the single most used isotope for all nuclear diagnosticimaging procedures-it accounts for a majority of the radiopharmaceutical market, and essentially encompasses the whole of the single-photon emission computed tomography (SPECT) modality in the low-energy field [1]. There are several production routes for generating 99m Tc [2,3], although the primary means of commercially derived 99m Tc is produced from 99 Mo (t 1/2 = 65.924 h; β − 100%) via neutron-driven nuclear reactions, of which, are dependent upon and characterised by the chemical and isotopic composition of the target material. For example, the use of either natural or enriched Mo or enriched U (20-90% 235 U) targets. Under these scenarios, the 98 Mo or 235 U targets are irradiated with a neutron source, such as a nuclear reactor, and form 99 Mo via 98 Mo(n,γ) 99 Mo or 235 U (n, fission) 99 Mo (~6% fission yield), respectively. The generated 99 Mo being sent to a processor is isolated, refined, and loaded onto a column, where Tc retention is negligible, such as alumina (Al 2 O 3 ) [4], and fixed in a portable generator for subsequent distribution. At the receiving end, whether it be a nuclear pharmacy or a hospital, the 99m Tc is eluted from the generator, compounded, and delivered to the point of use where it is administered to a patient. However, during this process a significant quantity of 99 Mo/ 99m Tc produced in this manner is typically lost due to decay (~1% per hour); the drug delivered at the

Production of 99 Mo, 101 Mo, 99m Tc and 101 Tc with a Neutron Generator
The liquid ammonium heptamolybdate (AHM) sample was irradiated with neutrons produced by an Adelphi DD110M neutron generator. Despite the high neutron generator yield of 2.2 × 10 10 n/s the thermal flux measured by gold foil activation was only 1.5 × 10 4 n/cm 2 ·s due to incomplete thermalization. The yield was determined from Bonner ball measurements that were consistent with the known yield from the generator arising from a 160 kV and D+ current of approximately 30 mA. Following irradiation, each sample was transferred to a high purity germanium (HPGe) detector for subsequent counting as shown diagrammatically in Figure 1a in order to detect for the presence of the short-lived species 101 Mo (t 1/2 = 14.61 min) and 101 Tc in the target material. Measurements were carried out for 10 min. The shielded HPGe detector is shown in Figure 1b. transferred to a high purity germanium (HPGe) detector for subsequent counting as shown diagrammatically in Figure 1a in order to detect for the presence of the short-lived species 101 Mo (t1/2 = 14.61 min) and 101 Tc in the target material. Measurements were carried out for 10 min. The shielded HPGe detector is shown in Figure 1b.  Figure 2 shows the gamma ray spectrum of AHM solution measured immediately following irradiation. Prominent signals from both the shorter-lived isotopes 101 Mo and 101 Tc are present directly following irradiation. Characteristic gamma lines for 101 Mo can be observed at ~590 keV (~19.2%), ~191 keV (18.2%), and 506 keV (11.6%), while gamma lines at ~307 keV (89.4%), ~545 keV (5.96%), and ~531 keV (1.00%) are indicative of 101 Tc. Much less discernible, but present in the spectrum, are the contributions arising from 99 Mo at ~740 keV (12.2%) and ~181 keV (6.1%). After the short irradiation and counting time relative to the half-life of 99 Mo, it would not be expected that significant 99m Tc would be present at this point, which is consistent with the data. The yields of 99 Mo, 101 Mo, and 101 Tc in the liquid target at the start of the counting measurement (t = 0) were calculated to be approximately 52 ± 3 Bq (1.3 ± 0.1 nCi), 1.2 ± 0.1 kBq (32.7 ± 3.5 nCi), and 4.6 ± 0.5 kBq (124.1 ± 13.2 nCi), respectively.  Figure 2 shows the gamma ray spectrum of AHM solution measured immediately following irradiation. Prominent signals from both the shorter-lived isotopes 101 Mo and 101 Tc are present directly following irradiation. Characteristic gamma lines for 101 Mo can be observed at~590 keV (~19.2%),~191 keV (18.2%), and 506 keV (11.6%), while gamma lines at~307 keV (89.4%),~545 keV (5.96%), and~531 keV (1.00%) are indicative of 101 Tc. Much less discernible, but present in the spectrum, are the contributions arising from 99 Mo at~740 keV (12.2%) and~181 keV (6.1%). After the short irradiation and counting time relative to the half-life of 99 Mo, it would not be expected that significant 99m Tc would be present at this point, which is consistent with the data. The yields of 99 Mo, 101 Mo, and 101 Tc in the liquid target at the start of the counting measurement (t = 0) were calculated to be approximately 52 ± 3 Bq (1.3 ± 0.1 nCi), 1.2 ± 0.1 kBq (32.7 ± 3.5 nCi), and 4.6 ± 0.5 kBq (124.1 ± 13.2 nCi), respectively.
Measurements showing the decay of 99 Mo and subsequent growth of its daughter, 99m Tc are shown in Figure 3. The plots were prepared from a series of hour-long spectra. The graphs provide the peak area which was evaluated via numerical integration techniques [20] for the 778 keV line of 99 Mo and the 140 keV line of 9m Tc. The characteristic curve for 9m Tc production is a result of the interplay between the half-life of the parent, 99 Mo (t1/2 = 65.924 h) and that of the daughter, 99m Tc (t1/2 = 6.007 h;). It should be noted that 101 Tc (t1/2 = 14.22 min) and 101 Mo (t1/2 = 14.61 min;) are also produced, but have much shorter half-lives ( Figure 3). Theoretically, the maximum activity of 99m Tc obtained from 99 Mo decay after 22.89 h of growing time [21], which is consistent with the data shown in Figure 3b. The calculated theoretical peak production yield of 99m Tc from the decay of 52 Bq of 99 Mo at approximately this time is approximately 35 Bq (0.95 nCi). Measurements showing the decay of 99 Mo and subsequent growth of its daughter, 99m Tc are shown in Figure 3. The plots were prepared from a series of hour-long spectra. The graphs provide the peak area which was evaluated via numerical integration techniques [20] for the 778 keV line of 99 Mo and the 140 keV line of 9m Tc. The characteristic curve for 9m Tc production is a result of the interplay between the half-life of the parent, 99 Mo (t1/2 = 65.924 h) and that of the daughter, 99m Tc (t1/2 = 6.007 h;). It should be noted that 101 Tc (t1/2 = 14.22 min) and 101 Mo (t1/2 = 14.61 min;) are also produced, but have much shorter half-lives ( Figure 3). Theoretically, the maximum activity of 99m Tc obtained from 99 Mo decay after 22.89 h of growing time [21], which is consistent with the data shown in Figure  3b. The calculated theoretical peak production yield of 99m Tc from the decay of 52 Bq of 99 Mo at approximately this time is approximately 35 Bq (0.95 nCi). Additionally, the spectra of 99 Mo and 99m Tc were recorded over a longer period of time ( Figure 4). The decay plots were used to determine the decay constants for the 778 keV peak of 99 Mo and the 141 keV peak for 99m Tc. The calculated decay constants for  Measurements showing the decay of 99 Mo and subsequent growth of its daughter, 99m Tc are shown in Figure 3. The plots were prepared from a series of hour-long spectra. The graphs provide the peak area which was evaluated via numerical integration techniques [20] for the 778 keV line of 99 Mo and the 140 keV line of 9m Tc. The characteristic curve for 9m Tc production is a result of the interplay between the half-life of the parent, 99 Mo (t1/2 = 65.924 h) and that of the daughter, 99m Tc (t1/2 = 6.007 h;). It should be noted that 101 Tc (t1/2 = 14.22 min) and 101 Mo (t1/2 = 14.61 min;) are also produced, but have much shorter half-lives ( Figure 3). Theoretically, the maximum activity of 99m Tc obtained from 99 Mo decay after 22.89 h of growing time [21], which is consistent with the data shown in Figure  3b. The calculated theoretical peak production yield of 99m Tc from the decay of 52 Bq of 99 Mo at approximately this time is approximately 35 Bq (0.95 nCi). Additionally, the spectra of 99 Mo and 99m Tc were recorded over a longer period of time ( Figure 4). The decay plots were used to determine the decay constants for the 778 keV peak of 99 Mo and the 141 keV peak for 99m Tc. The calculated decay constants for Additionally, the spectra of 99 Mo and 99m Tc were recorded over a longer period of time ( Figure 4). The decay plots were used to determine the decay constants for the 778 keV peak of 99 Mo and the 141 keV peak for 99m Tc. The calculated decay constants for both peaks were approximately 2.75 days or 66 h, which is indicative of the transient equilibrium relationship between the mother-daughter radioisotope pair, where the decay rate is established by the half-life of the mother isotope, i.e., 99 Mo.
The data shown in Figures 2-4 detail the production of 99m Tc occurred as a result of irradiation by the neutron generator and also indicates optimal times for extracting the 99m Tc from solution.  The data shown in Figures 2-4 detail the production of 99m Tc occurred as a result of irradiation by the neutron generator and also indicates optimal times for extracting the 99m Tc from solution. Figure 2 shows the presence of both Tc (i.e., 101 Tc) and Mo (i.e., 101 Mo and 99 Mo) isotopes in the irradiated AHM solution present at EOB. The produced Tc, likely in the form of the pertechnetate anion [TcO4] − , was subsequently extracted from the bulk Mo in the irradiated AHM solution using column chromatography with AC as the column substrate [22]. Gamma ray spectra ( Figure 5) of the solution and the extracted sample demonstrate successful removal of 101 Tc and 99m Tc from the solution onto AC. It is noted that isolation of 101 Tc was near total, however, some peaks of residual 101 Mo were identified (i.e., 506 keV and 591 keV) in the first hour following extraction (Figure 5a), likely due to inadequate washing of the column. Furthermore, no readily distinguishable peaks above the background from 99 Mo, which was also present in solution, were identifiable, and the most prevalent species on the AC is 101 Tc. The extraction performed after one-day was free from any 99 Mo/ 101 Mo, and only 99m Tc was found present (Figure 5b).   The data shown in Figures 2-4 detail the production of 99m Tc occurred as a result of irradiation by the neutron generator and also indicates optimal times for extracting the 99m Tc from solution. Figure 2 shows the presence of both Tc (i.e., 101 Tc) and Mo (i.e., 101 Mo and 99 Mo) isotopes in the irradiated AHM solution present at EOB. The produced Tc, likely in the form of the pertechnetate anion [TcO4] − , was subsequently extracted from the bulk Mo in the irradiated AHM solution using column chromatography with AC as the column substrate [22]. Gamma ray spectra ( Figure 5) of the solution and the extracted sample demonstrate successful removal of 101 Tc and 99m Tc from the solution onto AC. It is noted that isolation of 101 Tc was near total, however, some peaks of residual 101 Mo were identified (i.e., 506 keV and 591 keV) in the first hour following extraction (Figure 5a), likely due to inadequate washing of the column. Furthermore, no readily distinguishable peaks above the background from 99 Mo, which was also present in solution, were identifiable, and the most prevalent species on the AC is 101 Tc. The extraction performed after one-day was free from any 99 Mo/ 101 Mo, and only 99m Tc was found present (Figure 5b).  Figure 2 shows the presence of both Tc (i.e., 101 Tc) and Mo (i.e., 101 Mo and 99 Mo) isotopes in the irradiated AHM solution present at EOB. The produced Tc, likely in the form of the pertechnetate anion [TcO 4 ] − , was subsequently extracted from the bulk Mo in the irradiated AHM solution using column chromatography with AC as the column substrate [22]. Gamma ray spectra ( Figure 5) of the solution and the extracted sample demonstrate successful removal of 101 Tc and 99m Tc from the solution onto AC. It is noted that isolation of 101 Tc was near total, however, some peaks of residual 101 Mo were identified (i.e., 506 keV and 591 keV) in the first hour following extraction (Figure 5a), likely due to inadequate washing of the column. Furthermore, no readily distinguishable peaks above the background from 99 Mo, which was also present in solution, were identifiable, and the most prevalent species on the AC is 101 Tc. The extraction performed after one-day was free from any 99 Mo/ 101 Mo, and only 99m Tc was found present (Figure 5b).

Production of 99 Mo/ 99m Tc and 101 Mo/ 101 Tc Using a Neutron Generator and Mo Targetry
The production of the radioisotope mother-daughter pairs 101 Mo/ 101 Tc and 99 Mo/ 99m Tc using a D-D neutron generator and a liquid Mo target was demonstrated. Whereas the use of a neutron generator for producing 99 Mo/ 99m Tc from Mo targets for the consideration of commercial application has been previously reported upon [23], it is not apparent that this has been the case for 101 Mo/ 101 Tc. Historically, the 101 Mo/ 101 Tc pair has served as a signature in neutron activation studies, and in the context of 99 Mo/ 99m Tc it has been considered an impurity coinciding from the use of Mo targets [24,25]. Similarly, 101 Tc has been reported for cyclotron production of 99m Tc, although only as an impurity generated via the 100 Mo(p, γ) reaction.
Interestingly, the neutron capture properties of 98 Mo and 100 Mo are somewhat comparable. For example, the thermal neutron capture cross-sections are 0.130 b and 0.199 b for 98 Mo and 100 Mo, respectively, meanwhile the resonance integral cross-sections are 6.70 b for 98 Mo and 3.76 b for 100 Mo. Employing fast neutron regimes, such is the case with a D-D generator, permits access through down-scattering interactions, using a light moderator or within the sample itself, into the resonance regions with higher capture probabilities. However, the capture cross-sections for both isotopes drastically fall off at energies above 1 MeV, so it is noted that some moderation is required. Although the targetry investigated here was rudimentary in design, the authors envision the implementation of larger target masses and geometries where the target surrounds the neutron generator at the point of neutron production, better facilitating the interactions previously discussed.
Another key component of the system is the absence of detection of 99m Tc at the time of measurement, i.e., promptly after EOB. As discussed, this results from the short irradiation time relative to the half-life of the parent isotope 99 Mo, where no significant quantities of 99m Tc would have been expected to be produced through the decay of 99 Mo. It is noted that peak activity of 99m Tc as a result of decay typically occurs just short of 23 h-a characteristic feature exploited in the operation of 99m Tc commercial generators in radiopharmacies for elution scheduling, where elutions are performed once or twice a day as the industry standard. Therefore, either longer irradiation or longer decay times would generate 99m Tc (and 99g Tc; t 1/2 = 211,100 y) in the system. Coincidentally, this phenomenon permits several options when using natural isotopic targets or varying ratios of mixed enriched targets of 98 Mo/ 100 Mo. 1) The isolation of either 101 Tc or 99m Tc, where short irradiations and decay periods are ideal for 101 Tc, and longer irradiation periods and/or decay periods for 99m Tc, or 2) the co-production of 101 Tc and 99m Tc by employing long irradiation times with short decay periods. In the scenario for sole production of 101 Tc, one drawback using targets containing 98 Mo would be the potential build-in of 99 Mo, thus target irradiation schedules would require sufficient decay periods to remove 99 Mo/ 99m Tc/ 99g Tc from the system, where any residual 99m Tc/ 99g Tc could be removed through chemical processing prior to the next irradiation. For the latter production option, the ratio of 101 Tc: 99m Tc could be tailored for specific applications through the combination of enrichment ratios of 98 Mo: 100 Mo and the duration of irradiation.
Using the extrapolated data, the estimated 99 Mo (and its daughter 99m Tc) production yield as a function of time for a D-D neutron generator outputting 2 × 10 10 n/s, where a Mo target situated around the generator consumes/captures all of the neutrons produced and contains minimised impurities (an isotopically pure 98 Mo system is assumed), has been shown in Table 1. It is noted, that for simplicity neutron capture probabilities (i.e., capture cross-sections) are considered negligible, and the number of atoms in the target becomes equivalent to the neutron output (n/s) when complete neutron consumption within the target is assumed. Therefore, using Equation (1), only the neutron output (n/s) and the buildup term [1 − e −λt ] for 99 Mo are considered.

Inventory of Neutron Generators Required for Worldwide Annual Production
With a worldwide consumption of 99m Tc in excess of 40,000,000 patient doses annually, where each dose is equivalent to 740 MBq, a significant amount of the parent isotope 99 Mo must be generated on a weekly basis in order to fulfil demand. This is especially true when factoring in decay following EOB from the reactor or irradiation centre. The use of a neutron generator provides the possibility of generating isotopes at the point of use, for example at a hospital, a temporary field location, oncologist offices, etc. [23]. By using 99m Tc from 99 Mo that is continuously being generated and decaying in real time with a neutron generator, there are less inefficiencies and decay losses due to avoiding unnecessary cooling periods, manipulations, and transportation.
From Table 1, the estimated amount of doses of 99m Tc using two different production models was considered. The first model is represented by the commercial industry standard, whereby a target generating 99 Mo is irradiated for a week, decoupled from the production source, loaded into a generator device, and the 99m Tc resulting from the 99 Mo decay is "milked" approximately every 24 h [26]. The second model assumes the neutron generator is continuously irradiating the target and only the end-product (i.e., 99m Tc) is decoupled from the source of production. Under these circumstances, when production in the target material approaches saturation, the amount of 99 Mo produced is in near steady-state. Thus, the daily amount of 99m Tc that can be removed from the system remains constant and does not decrease, as is the case for standard commercial 99 Mo/ 99m Tc generators. Comparison of the two models and the efficiency gains for continuous generator is shown in Table 2. Table 2. Comparison of 99m Tc dose output (740 MBq) for a neutron generator (2 × 10 10 n/s) for the standard commercial 99 Mo/ 99m Tc generator production model after 1-week of irradiation (A 0 = 16.6 GBq), whereby the target is decoupled from the source of production, versus continuous production at saturation (~5-half-lives; A saturation = 19.4 GBq), where only the 99m Tc is removed from the irradiated target. Assumptions do not account for 99m Tc at EOB, generator elution efficiency, nor decay losses due to processing.

Continuous Generator (Doses)
Efficiency Gain (%)  101 Tc, and 24 h is 4.00 half-lives of 99m Tc. Therefore, more daughter radioisotope is generated due to faster parent decay and less loss due to daughter decay for 101 Mo/ 101 Tc in a comparable system as the 99 Mo/ 99m Tc commercial generators. Likewise, approximately 1.2 h would be required for reaching saturation activity (i.e.,~5-half-lives) of 101 Mo/ 101 Tc during irradiation; most commercial irradiations for 99 Mo/ 99 Tc extend up to 7 days or about 50% of saturation activity for 99 Mo. Furthermore, the decay of 101 Mo is solely 101 Tc and the end-decay member of the A = 101 isobar chain is 101 Ru. Therefore, losses in sequential labelling of 101 Tc due to lowered specific activity with competition of the presence of other Tc isotopes, such as with 99m Tc/ 99g Tc during radiopharmaceutical tagging, is not incurred. It is mentioned that the presence of 99g Tc arises from both the decay of 99 Mo directly (branching decay~13.95%) and via the isomeric transition decay of 99m Tc. Shown in Table 3, the total amount of neutron generators required to supply 40,000,000 doses of 99m Tc per year (109,589 doses per day) is calculated as a function of doses produced per day using targets of either natural Mo or pure 98 Mo, and the generator neutron flux, e.g., 2 × 10 10 n/s. and 2 × 10 12 n/s, where the former represents the output capabilities of D-D generators, as tested here, and the latter of D-T. The equivalent of 101 Tc when directly substituted for 99m Tc, using a 1:10 ratio factor for 99m Tc to 101 Tc, in a system is also considered. As a point of comparison, the IAEA reported that there are over 1500 cyclotron facilities distributed around the world [27]. Aiming for this number with the neutron generator model described here, it can be assumed that neutron outputs of~10 10 n/s would be too low, even when employing enriched Mo targets, and an excessive number of neutron generators would be required for fulfilling the global daily requirement of 99m Tc. However, at neutron outputs of~10 12 n/s, meeting this figure is easily achieved. When switching to the 101 Tc system, even the lower neutron output and natural Mo target provide a comparable number of required generators to the IAEA figure for cyclotrons, whereas at 10 12 n/s only a few systems would be required regardless of target composition.

Financial Considerations for Neutron Generator Production
Neutrons can be generated by various means, such as those listed in Table 4. The D-D neutron generators used in this research have both the lowest yield and also lowest cost compared to other approaches presented. Commercially available D-D neutron generators with yields close to 5 × 10 10 n/s, such as those used in this work [28], have recently become commercially available for applications such as boron neutron capture therapy (BNCT), where they can be deployed as an assembly of multiple generators clustered around a patient [29]. Similarly, clusters of D-D generators may be the most cost-effective route to isotope production, especially when production is intended to be deployed close to point-of-use. All of the systems listed in Table 4 require a dedicated facility with sufficient shielding for safe operation. In the case of D-D neutron generators the cost of the facility can be comparable to the cost of the neutron source, primarily because of the relatively low cost of the neutron source. When higher yields are required, linacs, cyclotrons and even reactors may prove the more cost-effective approach, but for small systems, the D-D neutron generator approach may have some merit.

Separation of Tc from LSA Mo Using AC
The affinity and uptake of Tc, particularly as [TcO 4 ] − , with AC is a well-known phenomenon and has been reported in many studies over the span of several decades [38][39][40][41][42][43][44][45]. Although the mechanism of Tc sorption onto AC is considered a complex process [46], there have been several driving mechanisms identified, such as electrostatic interactions, physisorption, chemisorption/ion-exchange, etc., which have been said to dictate this behaviour [47]. The [TcO 4 ] − anion is considered a relatively large anion with a low charge density, a low degree of hydration [48], and exhibits chaotropic character in regards to the Hofmeister series; it is found as the predominant form of Tc under oxidising conditions and across a range of pHs in most non-complexing media. The uptake of Tc on AC occurs at all pHs, however, at acidic pHs below the zero-point-charge of AC, whereby surface species become protonated and the surface becomes positively charged, the kinetics of Tc sorption are significantly enhanced [47,49,50]. Adsorption values greater than 99% [TcO 4 ] − (~1 MBq 99m Tc) onto AC within 1-min have been reported in the pH range 2-3 where maximum distribution coefficient (K d ) values were measured [51]. The primary contributing surface species are said to be carboxylic, carbonyl, laconic, and phenolic groups, where R-C=O and R-C-OH moieties serve as potential binding sites for [TcO 4 ] − [47,49,50] Measured K d greater than~10 5 for Tc uptake onto AC under acidic pHs have been reported [41,47,51]. The solution chemistry of Mo is much more intricate than that of Tc, where Mo speciation is highly dependent upon pH due to protonation/condensation/oligomerisation reactions, and it is quite sensitive to its chemical environment [53,54]. This behaviour is particularly apparent in acidic pHs where an assortment of anionic polymolydates with varying degrees of hydration and charge densities, and even cationic or neutral species, can (co)exist [53][54][55]. In solution, polyanionic Mo species, such as [MoO 4 ] 2− , are known to form hydrogen bonds stronger than in the surrounding bulk water, classifying them as cosmotropic in nature [56]. It has been shown that Mo adsorption on AC under static conditions is pH dependent, and once again is dictated by the surface charge of AC and the dominant species of Mo in solution [57][58][59][60]. At low pHs (~2-6), whereby the AC surface is positively charged and Mo exists as anionic species in solution (e.g., [ ), the interaction of Mo with AC is electrostatically favoured; AC surface functionalities described in the previous paragraph also apply here. Above this region (pH > 6), the AC surface is negatively charged, resulting from the disassociation of weakly acidic sites, and anionic Mo uptake is no longer favoured through electrostatic interactions [57,58]. The K d coefficients were determined using batch equilibrium techniques for the adsorption of [ 99 MoO 4 ] 2− onto AC with nitric acid concentrations ranging from 0.01 M to 2 M; it was found that K d values, slightly upwards of~10 3 , were highest at the lowest concentrations and dropped off rapidly with increasing nitric acid, likely due to the formation of non-anionic species [61].
The notion of 99m Tc adsorption on AC has been previously discussed in the literature [62]. In fact, the application of using AC to effectively separate 99m Tc from LSA 99 Mo, which was first demonstrated by Tatenuma et al. [63][64][65], has also been the topic of several subsequent studies from other groups [66][67][68]. For example, Tatenuma et al. demonstrated the separation of 99m Tc from Mo targets with specific activities <0.5 Ci/g of 99 Mo at neutral pH; isolation of [ 99m TcO 4 ] − in saline solution with a radiochemical purity of 6-7 N within 30-50 min and recoveries between 90-98% were achieved [63]. Production quantities of 99m Tc from µCi to hundreds of Cis per batch were determined feasible, and subsequent tagging experiments with HMPAO and MIBI proved comparable to fission-generator-derived 99m Tc radiopharmaceuticals in murine models [63].
Presented herein, a similar method of isolating Tc isotopes from irradiated LSA Mo solutions using AC has been demonstrated [22]; however, it is noted that separations were specifically performed under acidic conditions, where the kinetically-driven preferential uptake of Tc as [TcO 4 ] − over Mo, likely as anionic polyoxometalate species, onto AC could be exploited. Even at ultra-low specific activities of 101 Mo/ 99 Mo and for the shorterlived 101 Tc, effective removal of Tc from the solution was possible using a rudimentary separation platform.

Technetium-101 as a Potential Ephemeral Diagnostic, Therapeutic, and/or Theranostic Agent
In the field of nuclear medicine, theranostic isotopes are defined as a single isotope or a dual isotope pair, usually of the same element, that exhibits decay characteristics for both therapeutic (i.e., alpha, beta minus, Auger, etc.) and diagnostic decay modes (i.e., beta plus/annihilation, gamma. etc.). This category of isotopes may allow health practitioners to perform radiotherapeutic procedures, for instance for thyroid ablation or in oncological settings where other standard care treatments are not ideal, meanwhile imaging the radiation payload delivered to the site of interest. This may provide practitioners the opportunity to track the progress and success of a procedure as well as ensuring them that unnecessary dose to the patient is not incurred [69]. In contrast to procedures that employ isotopes of homologous elements for the imaging and therapy portions, e.g., 68 Ga/ 177 Lu, the use of single element theranostic systems has the added advantage and assurance that chemical and biochemical behaviour especially in vivo will not differ from each other [70].
The inventory of known technetium isotopes, characterised by masses from A = 85 to A = 120, are radioactive [71]. The hallmark isotope, most commonly used for diagnostic imaging on SPECT systems, for technetium is 99m Tc. Although the application of 99m Tc in Auger electron therapy has been proposed, other isotopes with more favourable nuclear properties are preferred [72], thus its use as a radiotherapeutic is hardly considered. Likewise, other Tc isotopes 94m Tc [73], 95g Tc and 96g Tc [74] have been identified for diagnostic uses in the PET modality, offering the possibility for acquisition of better resolution data or variation in tracer experiment durations.
However, it is not readily obvious in the literature that any Tc isotope has been considered or is currently used for therapeutic purposes, outside of the previously mentioned example of Auger therapy for 99m Tc. As an alternative, it is typical to substitute either 186 Re or 188 Re as a beta-minus emitting radiotherapeutic for 99m Tc, being that rhenium is the heavier chemical congener of Tc [75,76]. Although, Re and Tc share many chemical similarities, it is well-known that some pertinent physicochemical properties, such as redox potentials and predominant oxidation states, do differ [77], making the exchange of one for the other not so straightforward [78]. Even in simple systems, such as the binary halides, divergence between the chemistry of the two elements has often been observed [79]. Thus, the availability of a radiotherapeutic/theranostic isotope of Tc would serve as a powerful tool and provide better certainty under these scenarios. Ultimately though, the lack of a previously considered option for Tc likely stems from the fact that most of the accessible beta-minus emitting isotopes of Tc are either considered too long-lived, such as 98 Tc (t 1/2 = 4.20 × 10 6 y) or 99g Tc (t 1/2 = 2.11 × 10 5 y), or too short-lived, for example 100 Tc (t 1/2 = 15.56 s), for this application.
The general consensus amongst the scientific and medical communities has been that radiotherapeutic isotopes with half-lives ranging in between 6 h to 7 days were usually the most practical and well-suited as discussed by Qaim [80]. Although, Qaim also points out the biological half-life of an isotope is just as important, where consideration of both these parameters biological and physical half-lives constitute effective half-life. Other important factors relate to the pharmacokinetics of the radioisotope, such as the amount of time required for tagging, administration, site accumulation, and site retention, all of which are considered for determining and creating radiation dosimetry models [81]. However, the main argument against shorter-lived radioisotopes manifests from the feasibility and ease of: (1) delivery to the site of use and (2) storage of the isotope/drug.
Expanding interests in therapeutic and theranostic isotopes has presented a host of candidate radioisotopes, including ones that do not necessarily abide by the major criteria outlined by Qaim. A few such examples of these are 226 Th (t 1/2 = 30.57 min), which decays by multiple successive alpha emissions to long-lived 210 Pb [82], and the beta-minus emitter 214 Pb (t 1/2 = 27.06 min) and its succession of beta and alpha emitting daughter isotopes [83]. For both radionuclides, integration of the parent isotope, i.e., 230 U (t 1/2 = 20.23 d) for 226 Th and 222 Rn (t 1/2 = 3.82 d) for 214 Pb, into a generator form is a key feature that makes dispensation, tagging, and administering of the end drug possible despite the shorter than usual half-lives [84].
For commercial PET imaging, isotopes with half-lives on the order of minutes to seconds are often implemented, most of which are produced with an accelerator, i.e., cyclotron, for regional distribution and supply, or even directly at the site of use, e.g., hospital. Some of the most notable PET isotopes in this category include 15 O (t 1/2 = 122.24 s), 13 N (t 1/2 = 9.97 min), and 11 C (t 1/2 = 20.36 min). The use of automated synthesis modules and other streamlined manipulation techniques ease the production process, and despite their short-half lives, manufacturing is undertaken in accordance with strict good manufacturing practices (GMP) alongside QA-QC testing prior to dispensation.
By analogy, the suggested production and use for 101 Tc is likened to a combination of the aforementioned examples of accelerator-produced, short-lived commercial PET isotopes and the therapeutic radioisotopes 226 Th and 214 Pb. Like 226 Th and 214 Pb, 101 Tc is a relatively short-lived isotope and emits fundamental nuclear particles that are within the applicable range for therapeutic and diagnostic procedures. Unlike these isotopes, the parent isotope 101 Mo of 101 Tc does not allow for a generator-type scenario of usage. Likewise, this behaviour deviates from the production of PET isotopes, which are typically generated directly via transmutation with the charged particle beam, where no intermediate mother radionuclide precedes the end-product. However, as demonstrated here, 101 Tc can be produced straightforwardly through neutron capture reactions on 100 Mo to form 101 Mo followed by beta-minus decay. The shorter half-lives of both the parent and daughter permit ample activity of 101 Tc to be produced, even with moderate neutron fluxes and irradiation times under the conditions previously discussed, which in turn would allow for further tagging, QA-QC, and distribution within an immediate locale. Recent efforts on establishing robust and efficient separations of 99m Tc from low-specific activity 99 Mo targets [2,22,23,63,85,86] also provide possible avenues for the isolation and use of 101 Tc. As demonstrated here, the use of activated carbon as a chromatographic substrate is effective for isolating 101 Tc from LSA Mo solutions under acidic conditions. When linked to a production source, such as a neutron generator, this particular separation would allow continuous, streamlined and on-demand production and isolation.
As shown in Table 4, 101 Tc is a medium energy beta-minus emitter occurring at several prominent energies with the one at~487 keV (E βmax = 1320 keV) being the most distinct (~90%). In addition, there are a number of characteristic gamma emissions with the those at 306.8 keV (~89%) and 545 keV (5.9%) being the most notable. The final decay product of the A = 101 isobar chain and daughter product of 101 Tc is stable 101Ru. This is particularly advantageous considering that no long-lived radioisotopes are generated, such is the case with its sister isotope 99m Tc that yields 99g Tc, a byproduct often discarded into the environment post-imaging. Although its commercial use has never been proposed, particularly within the medical field, it can be argued that it would have a logical fit within the known toolbox of medical radioisotopes.
In comparison with other current commercially implemented therapeutic/theranostic beta-emitting radioisotopes (Table 5) [87], with the exception of half-life, 101 Tc exhibits unique, yet consistent decay modes. In fact, the most prominent beta energy at 487 keV is situated between theranostic radioisotopes 131 I and 177 Lu on the lower end, and the pure beta-emitters 89 Sr and 90 Y on the upper end. Interestingly, 101 Tc also lies between both radioisotopes of its elemental congener 186 Re and 188 Re. However, the primary beta of 89 Sr is closest in energy of those radioisotopes listed to 101 Tc, and it has a beta penetration range in tissue of approximately 8 mm [87]. A clinical application of 89 Sr is targeted radioisotope therapy of osseous metastases; in this regard, it is noted that one of the most common clinical applications of 99m Tc diagnostics is for skeletal scintigraphy also known as the "bone scan" with 99m Tc-methylene diphosphonate (MDP) or 99m Tchydroxydiphosphonate (HDP).
In addition, the primary characteristic gamma decay of 101 Tc is nearly identical to that of 131 I at 364.48 keV, which is often used for SPECT imaging. Not only is imaging feasible with commercial SPECT cameras, but ongoing technology, such as Compton cameras also provide an avenue for imaging options of higher energy gammas [17]. It is noted that whole body, simultaneous dual isotopic imaging of 99m Tc and 131 I has been reported for the use of locating functioning bone metastases arising from residual differentiated thyroid carcinoma (DTC) following 131 I thyroid ablation [88,89].
Although the shorter half-life of 101 Tc may seem to be a disadvantage, there are potential benefits that it may provide. One scenario is by lowering extraneous dose to the patient and public. It is envisioned that multiple therapy sessions could be conducted, with the ability to monitor and adjust the amount of radiation administered, over the same duration observed in a single session for current day radiotherapy treatments with longer-lived isotopes. The shorter half-life would also eliminate the need for patient isolation over extended periods, as such can be the case when high doses of radiation with longer half-lives are administered. Furthermore, from a production and supply perspective, many longer-lived radiotherapeutic isotopes are generated using nuclear reactors with high neutron fluxes in order to achieve workable activities and specific activities [90]. Accommodating for this can be logistically complex from the perspectives of scheduling and distribution, as well as financially expensive. On the other hand, the capacity for sufficient production using accelerators like neutron generators allows many of these issues to be circumvented. Because of the expansive and exhausted knowledge-base of 99m Tc and its utilisation in the field of nuclear medicine [91][92][93], it is not difficult for one to draw conclusions on what the potential possibilities and applications for 101 Tc could be. Being that it will ideally share nearly identical chemistry and chemical properties as those with 99m Tc, most tagging agents and kits would be translatable as well as the biochemical behaviour in vivo of the corresponding tagged moieties. Moreover, opportunities for dual isotopic imaging/therapies with 101 Tc and any of the aforementioned medically relevant Tc isotopes also becomes plausible. It is not within the scope of this study to pinpoint any particularly one purpose for 101 Tc as a diagnostic, therapeutic, or theranostic agent, but only to bring light to its prospective use.

Neutron Irradiation
The irradiated sample consisted of a solution of natural AHM (NH 4 ) 6 Mo 7 O 24 (50.

Gamma Spectroscopy
An Ortec trans-SPEC-DX-100P N-type HPGe detector was used ( Figure 1b) and had been calibrated using NIST traceable calibrated sources. The calibration sources were placed on top of the detector in the same location as the samples to be measured. The HPGe crystal was surrounded by lead shielding to reduce the background signals from the environment. The radioisotope 40 K is detectable at 1460 keV despite shielding. Using the data from the calibration, the measured efficiencies for the isotopes of interest were determined and are presented in Table 6.

Measurement of the Neutron Flux (Φ) Based on Gold (Au) Reference
A thin Au reference sample (99.99% purity) with dimensions of 8 mm × 15 mm and a mass of 0.5 g was irradiated for approximately 33.4 min. Activity measurements of the characteristic 411.80 keV peak (95.62%) of 198 Au (t 1/2 = 2.964 d) were made 18.5 h after the irradiation and were counted as 261 s.
When a sample is irradiated with flux, Φ thermal neutrons/cm 2 ·s, it becomes activated; the activity, in counts per second, A, is given by Equation (1): where N is the number of atoms in the Au reference sample, σ thermal is the neutron absorption cross-section of 197 Au (98.6 b) and t irradiation is the irradiation time. The mean lifetime λ relates to the half-life, t 1 2 of the element via Equation (2): Therefore, Φ can be calculated from the physical measurement of A from gamma spectroscopy using Equation (3): The integral of activity (I) between the times t 1 and t 2 from the end of the irradiation (t = 0) is given by Equation (4):

Extraction of Tc from Mo in a Bulk Sample of Irradiated AHM Solution
The extraction of Tc isotopes from bulk Mo in an irradiated AHM solution was performed using a column of activated carbon (~1 g). The AHM solution was passed over the column, washed with approximately 1 bed volume of the background solvent, and the column containing sequestered Tc was isolated for subsequent measurement by gamma spectroscopy.

Conclusions
In this paper, we demonstrate the production of 99 Mo/ 99m Tc and 101 Mo/ 101 Tc via neutron bombardment of an aqueous target of natural Mo. Although only relatively small quantities of these particular isotopes were produced under the rudimentary conditions tested, it is believed that improvements in experimental parameters, such as neutron flux, neutron moderation, target mass, target geometry, irradiation duration, extraction conditions, etc., would vastly enhance yields. The quantity of 99m Tc and 101 Tc using neutron generators over an infinite irradiation and isolation period has been extrapolated from this initial irradiation data set. With these data sets, the total number of neutron generators required to meet worldwide annual 99m Tc need using a highly mobile, exportable, and nonfission-based system was estimated and the financial implications were presented. Because of the potential efficiency gain in using 101 Tc for the same or similar imaging modalities and/or theranostics, the wide use of these systems is believed to be advantageous for the purpose of producing 99m Tc or 101 Tc without uranium, a nuclear reactor, or some more expensive, energy intensive, and less mobile accelerator such as a LINAC or cyclotron.

Patents
The work presented in this manuscript was validation for the patent under review: Mausolf, E. and Johnstone, E. (US/International Patent) Direct, Continuous Transmutation of Molybdenum (Mo) for the Production and Recovery of Technetium (Tc) and Ruthenium (Ru) using a Neutron Source, submitted 2018.