The availability of a sufficient ⁹⁹Mo supply involves a sequence of connected steps with high complexity. These steps can be defined as the ⁹⁹Mo supply chain. In the first step, uranium targets are fabricated and tested. Then, these targets are shipped to the irradiation facilities where the production of ⁹⁹Mo takes place. After that, the irradiated targets are transported to well-equipped processing facilities comprising hot cells, where the separation and purification of ⁹⁹Mo from other fission products are accomplished. Afterward, the purified carrier-free ⁹⁹Mo solutions are eventually used for the ⁹⁹Mo/^99mTc generator assembly process. Subsequently, these generators are globally distributed to provide ^99mTc of high quality, ready to use for different medical needs (NAP, 2018, 2016).

Based on the fact that ⁹⁹Mo decays with a relatively short half-life (T_1/2 = 65.94 h), all previous steps need to be achieved as rapidly as possible to minimize radioactive decay losses. The quantity of ⁹⁹Mo produced is defined with the term “six-day Curie”. This term represents how much ⁹⁹Mo radioactivity remains 6 days after the end of processing. Figure 3 highlights the six-day Curie concept through the ⁹⁹Mo supply chain (Paterson et al., 2015). The “six-day Curie” concept has been introduced by the manufacturers as a base for calibrating the sales price. However, it is not very suitable to express the losses of ⁹⁹Mo from the end of bombardment to arrival at the end user since neither transport to the processing facility, processing time, chemical yield, and transport times to the end user are accounted for.

Close to 10,000 uranium targets (among them 8,000 targets involving HEU) are consumed annually to afford a ⁹⁹Mo supply for nuclear medicine utilization. There are, at present, only four agencies that can manufacture and supply targets to the irradiation facilities. These laboratories are operated under the authority of the Atomic Energy Commission of Argentina (CNEA), Argentina; Canadian Nuclear Laboratories (CNL), Canada; Company for the Study and Production of Atomic Fuels (CERCA), France; and Nuclear Technology Products Radioisotopes (NTP), South Africa (NAP, 2018, 2016).

The neutron bombardment of the targets can be carried out in research reactors of high neutron flux, which is usually in the range of ≥ 10¹⁴ n/cm²s. Table 4 illustrates the potential irradiation reactors and their production volume (Zhuikov, 2014; NAP, 2018; NEA, 2019). The ⁹⁹Mo production capability of these reactors covers more than 95% of the global medical need for ⁹⁹Mo. The target irradiation process typically lasts five to 7 days until the ⁹⁹Mo growth reaches 80% of the saturation yield. At this point, the overall ⁹⁹Mo activity remains unchanged. That is because the quantity of ⁹⁹Mo produced by the target irradiation equals the loss of ⁹⁹Mo activity due to the radioactive decay (Paterson et al., 2015; NEA, 2017).

After the irradiation step, the targets are left to cool down, allowing short-lived and ultra-short-lived fission products to decay. Then, the irradiated targets are moved to the processing facilities where the recovery of ⁹⁹Mo from other fission products is conducted, and a pure ⁹⁹Mo-molybdate solution is prepared. Currently, there exist only five large-scale processing centers that can perform this task. Table 5 shows the five main processing facilities and their global ⁹⁹Mo supply (IAEA, 2013a, 2015; NAP, 2016,; NEA, 2019).

The uranium fission method deserves careful attention owing to its capability to provide adequate amounts of ⁹⁹Mo in a carrier-free form with high specific activity, which can satisfy the global medical need. However, the application of this technology faces inherent critical challenges. For instance, proliferation concerns are dramatically growing owing to the use of massive amounts of HEU targets. Nearly 80% of the global demand of ⁹⁹Mo is produced using uranium targets containing up to 93% of enriched ²³⁵U. During the production steps, about 50 kg of weapons-grade HEU are usually handled, and only a very minute amount (∼3%) is utilized in the entire process (NAP, 2009; NEA, 2010; IAEA, 2015). In order to eliminate these nuclear proliferation issues, HEU targets have to be fully substituted by LEU or natural uranium targets (Wu et al., 1995; Cols et al., 2000; Conner et al., 2000). In addition, the aging fleet of the currently used irradiation facilities is also of concern. Approximately 95% of the global supply of ⁹⁹Mo depends on only six reactors. With the exclusion of OPAL, all the other five reactors have been in service for more than 50 years. Lately, a number of unplanned shutdowns have taken place, and consequently, the international medical community has severely suffered from a ⁹⁹Mo shortage (Gumiela, 2018). In case of any unexpected future interruption, a renewed ⁹⁹Mo supply crisis cannot be ruled out and could seriously affect many countries due to the expected rise in the cost of dose per patient (NEA, 2011).

Moreover, the separation and purification of ⁹⁹Mo from fission products is a sophisticated process and calls for large and complex infrastructures, professional technical skills, and well-equipped laboratories. Only few centers worldwide are well-prepared to conduct this task (Ali and Ache, 1987; NAP, 2016, 2018; NEA, 2019). Furthermore, the generation of massive quantities of toxic radioactive waste containing medium and long-lived radionuclides has to be considered. A significant number of Curies (∼50 Ci) of fission radioactive waste are generated per production of one Curie (Ci) of ⁹⁹Mo (Boyd, 1987; IAEA, 1998; Tur, 2000). The difficulties mentioned above are clearly reflected in the continuing need for long-term capital investments and routine operating expenditures. As a result, the production cost of 1 Ci of ⁹⁹Mo from the fission route is four times higher than that of 1 Ci of neutron-capture-produced ⁹⁹Mo (Boyd, 1982a).

Sublimation is a physical phenomenon, which can be defined as the direct conversion of solids into a gas without moving through the liquid state. Perrier and Segrè were the first who described the possibility of a ^99mTc thermal extraction from ⁹⁹Mo based on the difference in the sublimation temperatures or volatilities of their oxides (Perrier and Segrè, 1937). Tc₂O₇ and MoO₃ become volatile at temperatures of 550 and 1,000°C, respectively (Perrier and Segrè, 1937; Childs, 2017). The separation steps were reported through several research studies. The irradiated ⁹⁹MoO₃ target is carefully placed inside a muffle furnace, in which a flow of oxygen stream is allowed to pass through. Then, the temperature is gradually elevated to about 800 °C. On heating, Tc₂O₇ is formed, which sublimates at 550°C. After that, the volatile Tc₂O₇ vapors are carried by the oxygen stream and trapped using a cooled surface. Finally, they are allowed to be completely dissolved in water or saline solution to obtain ^99mTcO₄ ⁻ (Hallaba and El-Asrag, 1975; Molinski, 1982; Zsinka, 1987; Miller, 1995; Christian et al., 2000; IAEA, 2013b).

This generator strategy generates a small volume of radioactive waste. In addition, it provides high specific activity ^99mTc even if ⁹⁹Mo of low specific activity origin is available. However, attempts to implement this technique in nuclear medicine departments as a clinical-based generator system suffered from several critical constraints. For instance, it includes the manipulation of bulk, complex, and fragile equipment, which needs a meticulous operational methodology and an advanced level of precautions and safety standards. Moreover, the ^99mTc separation yield is very low. Furthermore, it uses high heating temperatures, which may lead to a serious radiation contamination risk in case of any operational failure. Accordingly, these potential drawbacks discouraged further development plans in this direction.

The solvent extraction-based generator is a multi-step separation technique by which ^99mTcO₄ ^- can be recovered from an aqueous solution that contains the ⁹⁹Mo/^99mTc mixture by the extraction into an organic solvent (IAEA, 2013b). The separation of ^99mTc from its parent with the help of organic extractants was originally demonstrated by Gerlit in 1956 (Gerlit, 1956; Anders, 1960; Lebowitz and Richards, 1974; Molinski, 1982).

The separation process involves the dissolution of the irradiated ⁹⁹Mo target in an alkaline solution. Then, an organic solvent is added, which is followed by multiple-extraction cycles. Each cycle includes a vigorous shaking of the two immiscible phases to allow the ^99mTcO₄ ⁻ to migrate from the original aqueous solution to the organic extractant phase leaving its parent behind. After that, the extractant is segregated, washed several times, and thermally evaporated. Finally, the residual precipitate is redissolved in a saline solution. The process efficiency strongly depends on the selective capability of the extractant solvent for the ⁹⁹Mo/^99mTc pair. In other words, the process relies on the relative distribution or solubility ratio of the two radionuclides between the extractor and the original aqueous solution (Chattopadhyay et al., 2010, 2014; IAEA, 2013b; Mitra et al., 2020). D M 99 o   = ( C M 99 o ) e x t r a c t a n t ( C M 99 o ) a q u e o u s (19) D T 99 m c   = ( C T 99 m c ) e x t r a c t a n t ( C T 99 m c ) a q u e o u s (20)

Therefore, the ideal extractant should show a considerable selective affinity towards ^99mTc and neglected distribution selectivity for ⁹⁹Mo. The first used organic extractant to achieve the extraction process was Methyl Ethyl Ketone (MEK). After that, various organic solvents were reported (Allen, 1965; Anwar et al., 1968; Baker, 1971; Tachimori et al., 1971; Iqbal and Ejaz, 1974; Noronha, 1986; Bhatia and Turel, 1989; Bhatia and Turel, 1989; Minh and Lengyel, 1989; Taskaev et al., 1995; Chen and Tomasberger, 2001; Zykov et al., 2001; Skuridin and Chibisov, 2010). Even so, MEK is still considered the predominant extractant of choice because of its highly selective behavior for ^99mTcO₄ ^-, which permits the production of ^99mTc with high specific activity (IAEA 2017).

Practically, the solvent extraction generators are associated with a number of serious drawbacks that impede their wide-scale utilization; it is a complicated multi-step process with low extraction efficiency and requires a long separation time, resulting in the loss of ^99mTc radioactivity. Moreover, the separation step consumes large volumes of the organic extractant, creating significant radioactive waste. In addition, a strong possibility of fire hazards that result from the flammable nature of MEK. Furthermore, the poor radiation stability of organic solvents may heighten the contamination risk of the final ^99mTc product with some organic impurities.

The electrochemical-based ^99mTc generator can be introduced as a separation technique built on an oxidation-reduction reaction. This reaction is mainly governed by the difference in the formal reduction potential of the ⁹⁹Mo/^99mTc pair. Herein, the generator system consists of an electrochemical cell, in which the ^99mTc radionuclide can be separated from the ⁹⁹Mo/^99mTc equilibrium mixture by a reduction reaction on the surface of an electrode under the effect of an external voltage in accordance with its electrochemical reduction potential, i.e., electrochemical deposition. The ⁹⁹Mo/^99mTc pair exhibits different electrochemical reduction potentials, and their electrochemical reduction reactions can be illustrated as follows (Chakravarty et al., 2012b; 2012c, 2011, and 2010a): M o O 4 2 − + 4   H 2 O + 6 e −   → M o   + 8   O H −   E o = − 1.05   V (21) T c O 4 − + 4   H + + 3   e −   → T c O 2   + 2   H 2 O   E o = + 0.738   V (22)

It can be observed that the ^99mTcO₄ ⁻ species have a higher electrochemical reduction potential than the ⁹⁹MoO₄ ^2- ions, which offers a positive advantage in facilitating the separation step.

The process starts with the dissolution of the irradiated ⁹⁹Mo target in a strong alkaline electrolyte (pH ∼ 13). Then, a constant potential is applied and adjusted for a certain period to permit the ^99mTcO₄ ^- species to be electro-deposited on the surface of the cathode electrode. After that, this cathode is removed from the working cell and transferred to a new cell, where the deposited ^99mTc can be stripped back and redissolved in a saline solution by applying a high reverse potential (reverse electrode polarity) for a short period. By this technique, ^99mTc can be oxidized back to the solution in the form of ^99mTcO₄ ^-. Eventually, ^99mTc eluate is allowed to pass through an alumina column to minimize the ⁹⁹Mo contamination level.

In order to achieve a satisfactory ^99mTc separation yield, some parameters have to be well-optimized during the electrolysis process. For instance, careful attention should be devoted to adjusting the effective potential, proper selection of suitable electrodes and electrolysis medium candidates, and successful control of the electrolyte pH and separation time.

The applied potential is the crucial factor for a successful extraction. Generally, the electrochemical separation can be performed by either applying a constant current (galvanostatic) or a constant potential (potentiostatic) conditions. However, the potentiostatic condition offers favorable ^99mTc deposition and, in the meantime, limits the precipitation of ⁹⁹Mo and/or any other contaminants. The electrolysis potential should be adjusted to be more negative than the formal reduction potential of ^99mTcO₄ ^- species and less negative than the electrochemical reduction potential of ⁹⁹MoO₄ ^2- ions. Therefore, the optimum reduction potential for the ^99mTc generator system should be in the region of: − 1.05   V > e l e c t r o l y s i s   p o t e n t i a l   >   + 0.738   V

Additionally, it needs to be fully compatible with the used electrolyte to avoid any possibility of chemical degradation during the separation process. Several classes of electrolytes of organic or aqueous origin can be used. The aqueous electrolytes show higher radiation stability during the course of the electrolysis compared to organic liquids, which results in fewer chemical impurities associated with the produced ^99mTc radioactive solution. However, their use is usually associated with the evolution of hydrogen gas, which lowers the pH of the electrolyte medium and may decrease the separation performance. For the working electrodes, they need to be chemically inert with considerable electrochemical and radiation resistance. The temperature of the medium should be adjusted sufficiently below the electrolyte boiling point to avoid its rapid evaporation, which results in unacceptable ^99mTc solution purity. The reaction time should be established to achieve the highest possible ^99mTc yield in the shortest period to minimize ^99mTc radioactivity loss (Chakravarty et al., 2012b; 2012c, 2011; 2010a; 2010b).

Based on this method, Chakravarty et al., 2010a, reported the production of ^99mTc from a ⁹⁹Mo/^99mTc generator. The daughter extraction was performed in an electrochemical cell by applying a constant potential of 5 V in sodium hydroxide electrolyte (pH ∼ 13). The electrolysis process lasted for approximately 1 hour. Hence, the ^99mTc selectively accumulated on the platinum electrode (cathode). The ^99mTc deposited was recovered in a 0.9% saline solution. Then, it was purified with the use of an alumina column (Chakravarty et al., 2010a).

The prospect of ⁹⁹Mo/^99mTc electrochemical-based generators provide an adequate ^99mTc yield in a non-carrier added form with a satisfactory purity level for radiopharmaceutical applications. In addition, their production capacity neither relies on the amount of the used sorbent, nor the extractant volume. Nevertheless, this production strategy is implicated with some opposing weaknesses that preclude its practical medical use. For example, the electrolysis step is accompanied by the evolution of oxygen and hydrogen gases, which may pose an explosion hazard. In addition, the oxygen gas may inhibit the reduction of ^99mTcO₄ ^- species, which drastically decreases the separation yield. Moreover, the working electrodes require tedious cleaning processes directly after each separation cycle. Furthermore, it is a relatively time-consuming production technique, including different consecutive separation and purification steps, which consequently result in a marked loss of ^99mTc radioactivity. Finally, the procedure needs high shielding requirements to reduce the exposure dose to the working personnel. For these reasons, this strategy strongly requires a very rigorous operating regime with a high degree of safety and precautionary measures and well-trained operators with high technical awareness of electro- and radiochemistry fundamentals.

This generator system can be introduced as a modification or an updated version of the solvent extraction method (Dhami et al., 2007; Dutta and Mohapatra, 2013; Chang, 2016; Davarkhah et al., 2018). It involves the separation of ^99mTc species from ⁹⁹Mo with the use of a membrane as a semi-stationary phase and two other mobile phases that are located on both sides. These three compartments can be described as follows (Sastre et al., 1998; Chaudry, 2000; Yassine, 2000):

1. The primary aqueous solution (the feeding phase): It consists of the ⁹⁹Mo/^99mTc mixture.

The extraction process is performed in three simultaneous steps. In the first step, the ^99mTc species are selectively collected from the feed mixture by the extractant. Then, they travel through the membrane to the other side, where they are eventually exchanged with another species from the clean receiving solution. This journey of ^99mTc from the primary solution to the membrane and from the membrane to the receiving solution is only controlled by separation reaction kinetics without the help of any external driving force. Therefore, it is only attributed to the concentration difference on both sides of the membrane.

Many different supported liquid membrane frameworks have been studied to separate ^99mTc on the laboratory scale (Sastre et al., 1998; Gyves and Rodríguez, 1999; Chaudry, 2000; Yassine, 2000; Chen et al., 2001). Nonetheless, a number of inherent practical difficulties still ask for more research efforts. For instance, it is a time-consuming strategy and involves the loss of the majority of ^99mTc radioactivity attributable to the remarkably slow extraction kinetics. In addition, it supplies unsatisfactory ^99mTc separation yield with high ⁹⁹Mo breakthrough and organic chemical impurities due to the low radiation resistance of the organic liquids and limited life-course of the used membrane. Furthermore, it generates substantial amounts of radioactive waste.

Several ⁹⁹Mo/^99mTc generator production strategies have been explored and developed over the last few decades. Nevertheless, the column chromatography-based generator type has attracted considerable interest as a reliable source to supply ^99mTc ready for nuclear medicine investigations. As a result, the chromatographic column generator is considered the most predominant choice for the hospitals and nuclear medicine departments owing to many valuable benefits, including (Nawar and Türler, 2019):

• One-step process: The ^99mTc radioactivity can be extracted in only one step, effectively minimizing the radioactive decay losses.

In generic terms, this generator system consists of two phases, i.e., a stationary phase (the sorbent material) and an external liquid phase (the eluent solution). The central concept is built on the retention of ⁹⁹Mo radionuclide on a column incorporated with proper sorbent material. Then, the ⁹⁹Mo decays to yield ^99mTc, followed by the elution of the generated daughter with a suitable isotonic saline solution. The separation of the ^99mTc species is mainly based on the difference between their low retention affinity on the column material and their high solubility in the eluent solution. The parent ⁹⁹Mo exists in the molybdate form, M 99 oO 4 2 − , which in a weakly acidic medium, polymerizes to polymoybdate species, as follows (Dash and Chakravarty, 2016; Nawar and Türler, 2019): 7   M 99 o O 4 2 − + 8   H +   ⇌ M o 7 99 O 24 6 − + 4   H 2 O (23)

The formed species with six negative charges offer favorable conditions for interacting with the positively charged sorbent surface. After that, the polymoybdate M o 7 99 O 24 6 − decays to form the technetium pertechnetate (^99mTcO₄ ^-), which possesses only one negative charge. Hence, the binding ability of ^99mTcO₄ ^- becomes weaker than ⁹⁹Mo₇O₂₄ ^6- with the stationary phase (Dash and Chakravarty, 2016). Then, an isotonic saline solution is subsequently passed through the column to replace the weakly bound ^99mTcO₄ ^- species with Cl⁻ ions to produce sodium pertechnetate (Na^99mTcO₄), leaving the ⁹⁹Mo on the column, according to the equations: M o 7 99 O 24 6 − + 6   R +   → R 6 M o 7 99 O 24   → β −     R T 99 m c O 4 − + 5   R + (24) R T 99 m c O 4 − + N a C l   → N a T 99 m c O 4 + R C l (25)

The generator performance greatly depends on the sorption–elution kinetics and the degree of the sorbent selectivity for the parent Mo (containing ⁹⁹Mo). Therefore, the careful selection of the column material is the prime factor for the success of this generator system. Favorable sorbent material should have a considerable selectivity for Mo and, at the same time, a very limited or no selective capability for ^99mTc. Based on this parameter, an optimum separation environment can be established between the two radionuclides, and thereby, the ^99mTc can be extracted in high purity. Furthermore, fast elution kinetics of the daughter can be obtained.

Moreover, the sorbent material should show a high sorption capacity for Mo (containing ⁹⁹Mo). The sorption capacity indicates the quantity of Mo the column material can retain. It mainly depends on the number of available active sites on the sorbent surface for the parent element. Higher sorption capacity is directly reflected by a lesser amount of column material needed to achieve the desirable radioactivity level. Consequently, it is translated to a higher radioactive concentration of the eluted ^99mTc solution.

Furthermore, the column material needs to have a high radiation resistance and suitable chemical stability to withstand the intense radioactive irradiation and the chemical environments (Nawar and Türler, 2019). In this regard, varieties of column chromatographic generators have been developed. The differences among these generators are the sorbent material and the origin of ⁹⁹Mo. The following subsections highlight different column chromatographic generators approaches and their utilization to produce ^99mTc for medical use.

Following the delivery of the irradiated ⁹⁹Mo targets, a radioactive ⁹⁹Mo/^99mTc mixture is prepared by dissolving the irradiated targets in 2M NaOH according to the following reaction: M 99 o O 3   + 2 N a O H     →   N a 2 M 99 o O 4   +   H 2 O (26)

The formed Na₂ ⁹⁹MoO₄ solution is allowed to pass through a stainless steel column loaded with 5 g of AC, by which a particularly selective extraction of ^99mTc can take place. The AC column has high sorption efficiency towards ^99mTc compared to ⁹⁹Mo and other radio-contaminants present in the solution. The reason for this selective sorption behavior capability is still under investigation.

During the second step, small amounts of ⁹⁹Mo contaminants may be adsorbed on the activated charcoal column. In order to remove these impurities, three successive elution steps have to be performed. In the first step, the column is rinsed with a sufficient amount of water. Then, ⁹⁹Mo is eluted with sodium hydroxide solution. Finally, the column is rewashed with water to remove any residual quantities of ⁹⁹Mo and NaOH. The flushed ⁹⁹Mo solution can be allowed to decay to be reused again. This step may be repeated many times until the ⁹⁹Mo no longer has enough activity to be utilized.

In this step, the stainless steel column containing the activated carbon is heated up to 80–90°C. In the meantime, a sufficient amount of water is passed through the column. Hence, the ^99mTc radioactivity can be collected as a warm, weakly alkaline solution.

The obtained solution from the previous step is mixed with sodium chloride solution to form a 0.9% saline solution.

After completing the elution process (step 4), the ^99mTc solution may still contain some undesirable radio-impurities such as ^92mNb, ⁹⁵Nb, and ⁹⁶Nb, which arise from the neutron activation of stable radioisotopes of molybdenum, in addition to the parent ⁹⁹Mo. In order to purify the ^99mTc eluate from these radio-contaminants, additional purification steps have to be carried out. Therefore, the resulting ^99mTc solution from step 5 is allowed to pass through two successive columns; IER and AL columns, respectively.

The TcMM technology offers a new possibility for separating highly purified ^99mTc from ⁹⁹Mo of low and medium-specific activity. The produced ^99mTc shows high labeling efficacy with several pharmaceutical kits. Nevertheless, this technology still requires approval for making human-use ^99mTc radiopharmaceuticals. The TcMM can utilize ⁹⁹Mo radioactivity in the range of kBq to TBq and produce ^99mTc of a maximum of 500 Ci (18.5 TBq) per batch (Tatenuma et al., 2016a, 2014). However, the produced ^99mTc can only be distributed locally to the usage sites near the separation centers. In addition, the loss of some of the ^99mTc radioactivity during the multi-separation and purification steps and the transportation to more distant usage sites is a marked disadvantage.