Science and Technology of Nuclear Installations

Science and Technology of Nuclear Installations / 2014 / Article
Special Issue

Molybdenum-99 (99Mo): Past, Present, and Future

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Review Article | Open Access

Volume 2014 |Article ID 345252 | https://doi.org/10.1155/2014/345252

Van So Le, " Generator Development: Up-to-Date Recovery Technologies for Increasing the Effectiveness of Utilisation", Science and Technology of Nuclear Installations, vol. 2014, Article ID 345252, 41 pages, 2014. https://doi.org/10.1155/2014/345252

Generator Development: Up-to-Date Recovery Technologies for Increasing the Effectiveness of Utilisation

Academic Editor: Pablo Cristini
Received30 Jun 2013
Accepted05 Aug 2013
Published16 Jan 2014

Abstract

A review on the sources available today and on the generators developed up to date for increasing the effectiveness of utilisation is performed in the format of detailed description of the features and technical performance of the technological groups of the production and recovery. The latest results of the endeavour in this field are also surveyed in regard of the technical solution for overcoming the shortage of supply. The technological topics are grouped and discussed in a way to reflect the similarity in the technological process of each group. The following groups are included in this review which are high specific activity : the current issues of production, the efforts of more effective utilisation, and the high specific activity -based generator and concentration units; low specific activity : the production based on neutron capture and accelerators and the direct production of and the methods of increasing the specific activity of using Szilard-Chalmers reaction and high electric power isotopic separator; up-to-date technologies of recovery from low specific activity : the solvent extraction-based generator, the sublimation methods for / separation, the electrochemical method for recovery, and the column chromatographic methods for recovery. Besides the traditional -generator systems, the integrated generator systems ( generator column combined with postelution purification/concentration unit) are discussed with the format of process diagram and picture of real generator systems. These systems are the technetium selective sorbent column-based generators, the high Mo-loading capacity column-based integrated generator systems which include the saline-eluted generator systems, and the nonsaline aqueous and organic solvent eluent-eluted generator systems using high Mo-loading capacity molybdategel and recently developed sorbent columns. concentration methods used in the recovery from low specific activity are also discussed with detailed process diagrams which are surveyed in two groups for concentration from the saline and nonsaline -eluates. The evaluation methods for the performance of -recovery/concentration process and for the -elution capability versus Mo-loading capacity of generator column produced using low specific activity source are briefly reported. Together with the theoretical aspects of / and sorbent chemistry, these evaluation/assessment processes will be useful for any further development in the field of the recovery and / generator production.

1. Introduction

The development of the original generator was carried out by Walter Tucker and Margaret Greens as part of the isotope development program at Brookhaven National Laboratory in 1958 [1]. is currently used in 80–85% of diagnostic imaging procedures in nuclear medicine worldwide every year. This radioisotope is produced mainly from the generators via -particle decay of its parent nuclide . nuclide decays to with an efficiency of about 88.6% and the remaining 11.4% decays directly to . A generator, or colloquially a “technetium cow,” is a device used to extract the -pertechnetate generated from the radioactive decay of (). As such, it can be easily transported over long distances to radiopharmacies where its decay product () is extracted for daily use. sources used in different generators are of variable specific activity (SA) depending on the production methods applied. Based on the nuclear reaction data available today, two types of sources of significantly different SA values (low and high SA) can be achieved using different production ways. Accordingly, generators using low or high SA should be produced by suitable technologies to make them acceptable for nuclear medicine uses. The safe utilisation of the generators is definitely controlled by the quality factors required by the health authorities. However, the acceptability of the generator to be used in nuclear diagnostic procedures, the effective utilisation of generator, and the quality of -based SPECT imaging diagnosis are controlled by the generator operation/elution management, which is determined by the concentration of the eluate/solution. This also means that the efficacy of the generator used in nuclear medicine depends on the concentration of the solution eluted from the generator, because the volume of a given injection dose of -based radiopharmaceutical is limited. The current clinical applications of are shown in Table 1. As shown, the injection dose activity of -based radiopharmaceutical delivered in 1 mL solution is an important factor in determining the efficacy of the solution produced from the generators. So it is clear that the concentration of the solution eluted from the generator is the utmost important concern in the process of the generator development, irrespectively using either fission-based high specific activity or any source of low specific activity. It is realised that a complete review on the and production/development may contribute and stimulate the continuing efforts to understand the technological issues and find out the ways to produce a medically acceptable generator and to overcome the shortage/crisis of supply. So this review is to give a complete survey on the technological issues related to the production and development of high and low specific activity and to the up-to-day recovery technologies, which are carried out in many laboratories, for increasing the effectiveness of utilisation. The evaluation methods for the performance of the -recovery/concentration process and for the -elution capability versus Mo-loading capacity of the generator column produced using (or any low specific activity source) are briefly reported. Together with the theoretical aspects of and sorbent chemistry, these evaluation/assessment processes could be useful for any further development in the field of the recovery and generator production. The achievements gathered worldwide are extracted as the demonstrative examples of today progress in the field of common interest as well.


Organ 99mTc radiopharmaceuticalInjection activity dose (*) Organ 99mTc radiopharmaceuticalInjection activity dose (*)

Brain 99mTc-ECD10–20 mCi Kidney 99mTc-MAG35–15 mCi
99mTc-ceretec (HmPAO)10–2 mCi 99mTc-DTPA5–15 mCi
Lung 99mTc-MAA2–4 mCi 99mTc-Gluceptate5–15 mCi
99mTc-DTPA aerosol30 mCi/3 mL
(10 mCi/mL)
99mTc-DMSA2–5 mCi
99mTc-Technegas100–250 mCi/mL Skeleton 99mTc-MDP10–20 mCi
Thyroid 99mTc-pertechnetate5–10 mCi 99mTc-HDP10–20 mCi
Liver 99mTc-IDA5–10 mCi Heart 99mTc-Sestamibi10–30 mCi
99mTc-sulfur/albumine colloid5–15 mCi 99mTc-PYP10–15 mCi
Spleen 99mTc-sulfur/albumine colloid2-3 mCi 99mTc-Tetrofosmin5–25 mCi
99mTc-red blood cells2-3 mCiTumour 99mTc-Sestamibi15–20 mCi

2. High Specific Activity : Current Issues of Production and Efforts of More Effective Utilisation

2.1. Production of High Specific Activity

High SA is currently produced from the uranium fission. The fission cross-section for thermal fission of 235U is of approximately 600 barns. 37 barns of this amount result in the probability of a atom being created per each fission event. In essence, each one hundred fission events yields about six atoms of (6.1% fission yield). Presently, global demand for is met primarily by producing high specific activity (SA) from nuclear fission of 235U and using mainly five government-owned and funded research reactors (NRU, Canada; HFR, the Netherland; BR2, Belgium; Osiris, France; Safari, South Africa). After neutron bombardment of solid uranium targets in a heterogeneous research reactor, the target is dissolved in a suitable solution and the high SA is extracted, purified and packed in four industrial facilities (MDS Nordion, Canada; Covidien, the Netherland; IRE, Belgium; NTP, South Africa), and supplied to manufacturers of generators around the world [212]. CNEA/INVAP (Argentina), ANSTO (Australia), Russia, and BATAN (Indonesia) also produce fission and total supply capacity of these facilities is about 5% of the global demand of [3]. The weekly demand of is reported to be approximately 12000 Ci at the time of reference (6-day Ci). This is equivalent to 69300 Ci at the end of bombardment (EOB). All five of the major production reactors use highly enriched uranium (HEU) targets with the isotope 235U enriched to as much as 93% to produce (except Safari 1 in South Africa which uses 45% HEU). As mandated by the US Congress, non-HEU technologies for and production should be used as a Global Initiative to Combat Nuclear Terrorism (GICNT) [13, 14]. The production plans for conversion of HEU to low enriched uranium (LEU) based technology, using heterogeneous research reactors, achieved a major milestone in years 2002–2010 and currently the production of high SA from LEU targets is routinely performed in Argentina (from 2002), in Australia (from 2009), and in South Africa (from 2010). CNEA/INVAP (Argentina) is a pioneer in the conversion of HEU to LEU by starting LEU-based production in 2002 after decommissioning of HEU technology which has been operated 17 years ago [15, 16]. INVAP also demonstrated the maturity of LEU technology via technology transfer to ANSTO for a modest industrial scale manufacture of a capacity of 300–500 6-day curies per batch. With an announcement last year on a great expansion of production capacity of LEU-based facility being started in 2016 in Australia [17], ANSTO and CNEA/INVAP will become the first organisations confirming the sustained commercial large-scale production of based on LEU technology. High SA is of approximately 50,000 Ci /g of total Mo at EOB (The OPAL reactor, Australia, thermal neutron flux: 9.1013 n/cm−2 sec−1), irrespectively using either HEU or LEU-based fission technologies. With the effort in maintaining the supply of high SA , several alternative non-HEU technologies are being developed. Fission of 235U to produce is also performed using homogeneous (solution) nuclear reactor and recovery system, so-called Medical Isotope Production System (MIPS) [18]. The reactor fuel solution in the form of an LEU-based nitrate or sulphate salt dissolved in water and acid is also the target material for production. In essence, the reactor would be operated for the time required for the buildup of in the fuel solution. At the end of reactor operation, the fuel solution pumped through the -recovery columns, such as Termoxid 52, Termoxid 5M, titana, PZC sorbent, and alumina, which preferentially sorbs molybdenum [19, 20]. The is then recovered by eluting the recovery column and subsequently purified by one or more purification steps. It is estimated that a 200 kW MIPS is capable of producing about 10,000 Ci of at the end of bombardment (five-day irradiation) [2, 18, 21]. The possibility of using the high power linear accelerator-driven proton (150–500 MeV proton with up to 2 mA of beam current, ~1016 particles/s) to generate high intensities of thermal-energy neutrons for the fission of 235U in metallic LEU foil targets has been proposed [2, 22]. This accelerator can produce an order of magnitude more secondary neutrons inside the target from fission. The low energy accelerator (300 keV deuteron with 50 mA of beam current)-based neutron production via the D,T reaction for the fission of 235U in LEU solution targets has been reported [2]. The fission of 235U for the production can be performed with neutrons generated from the >2.224 MeV photon-induced breakup of D2O in a subcritical LEU solution target. Accelerator-driven photon-fission  238U(γ,f) is also proposed as an approach to produce high SA using natural uranium target [2, 2325].

Under the consultation for the fission plant in ANSTO, the author of this review paper has proposed a project of “Automated modular process for LEU-based production of fission  [26]. The consent of the Chief Executive Officer of ANSTO is a positive signal that might get scientists ahead of the game with next generation (cheaper, better, and faster) Mo-99 plant design. The aim of this project is to provide the integrated facility, composed of automated compact high technology modules, to establish medium-scale production capability in different nuclear centres running small reactors around the world. In essence, this project is to decentralize the production/supply and the radioactive waste treatment burden in the large facilities and to bring production closer to users ( generator manufacturers) to minimize the decay loss. The modular technology-based production is standardized for the secure operation sustainable with the supply of replaceable standardized modules/components for both processing and radioactive waste treatment. The above-mentioned objectives are in combination to solve basically the undersupply problem or crisis by increasing the numbers of smaller processing facilities in hundreds of nuclear centres owning production-capable reactors in the world and to reduce the cost of for patient use. The brief of the modular technology is the following. Currently, three main medical radioisotopes , 131I, and 133Xe are routinely produced from uranium fission. So, it is conceivable to say that the fission uranium based medical isotope production facility is composed of 6 main technological modules: target digestion module, separation module, 131I separation module, 133Xe separation module, uranium recovery module, and waste treatment modules (gas, solid, and liquid waste modules). For production alone, the numbers of main modules can be reduced to 4, comprising main module for uranium target digestion; main module for separation; main module for uranium recovery; main module for waste treatment (gas, solid, and liquid waste modules).

Each main module in this description is composed of several different functional modules. As an example, the main module for separation incorporates 7 functional modules, such as five ion exchange resin/sorption functional modules and two solution delivery functional modules (radioactive and nonradioactive).

A pictorial description of the structure of one main module which is capable of incorporating five functional modules (below illustrated with two functional modules as examples) is shown in Figure 1.

The operation of this main module is automated and computerized. The integrated fluid flow and radioactivity monitoring system using photo and/or radiation diode sensors provides the feedback information for safe and reliable process control. The in-cell maintenance based on the replacement of failed functional module is completed quickly ensuring continuous production run. Advantages of this facility setup are the following: compact system with controllable and reliable process; less space required that minimizes the cost of the facility (one double-compartment hot cell for whole process); minimal maintenance work required that due to highly standardized modular integration; high automation capability; low cost production of making this modular technology feasible for small nuclear research centres in many countries of the world; centralizing the module supply and maintenance giving high security and sustainability of production to small producers with few resources; high capability of the network-based production/supply to overcome any global crisis.

The W impurity in massive LEU targets is still challenging the quality of obtained from different recovery processes, because the ions and radioactive impurity (188Re) generated from neutron-activated W cause serious problems in the generator manufacture and in the use of -pertechnetate solution, respectively. The effort to remove W impurity from the solution produced from LEU target is being performed as shown in Figure 2 [27].

2.2. High Specific Activity Fission -Based Generators and Concentrators

The isolation of from uranium fission typically generates with a specific activity greater than >10,000 Ci/g at the six-day-Ci reference time (specific activity of carrier-free is 474,464.0 Ci/g [28]). This SA value permits extraction of the daughter nuclide using chromatographic alumina column [1, 2935]. Today, most commercial generators are designed by taking advantage of much stronger retaining of the anions compared with the anions on acidic alumina sorbent. Although the adsorption capacity of the alumina for anions is low (<10 mg Mo/g), the very low content of Mo in the high SA solution (0.1 mg Mo per Ci ), which is loaded on a typical column containing 2-3 g of alumina for a 4 Ci activity generator, ensures a minimal breakthrough in the medically useful -pertechnetate solution extracted from the generator system. When the decays it forms pertechnetate () which is easily eluted with saline solution from the alumina column resulting an injectable saline solution containing the in the form of sodium-pertechnetate. The most stable form of the radionuclide in aqueous solution is the tetraoxopertechnetate anion. The most important requirement for the design of an alumina column-based recovery system is that it must exhibit both a high elution efficiency (typically >85%) and minimal breakthrough (<0.015%) [36, 37]. The generators are sold on the world market with different sizes from 200 mCi to 4000 mCi and the elution of is performed with 5–10 mL normal saline. Fission -based generators commercially available in the US are of the activity range between 0.2 Ci and 4.0 Ci at the six-day curies reference time and in ANSTO (Australia) between 0.45 Ci to 3.2 Ci. The cost-effective utilisation of a generator and the quality of based single photon emission computed tomography (SPECT) imaging diagnoses is controlled by the generator operation/elution management. The primary factor pertaining to the nuclear medicine diagnostic scans’ quality is the concentration of obtained from the generator elution, which is expressed as activity per mL. The injection dose activity of -based radiopharmaceuticals delivered in 1 mL solution (-concentration, mCi/mL) is an important factor in determining the useful life time of the generators and the quality of based SPECT imaging diagnosis as well. Generally, a eluate is produced from the generator in fixed volume and the concentration of the in the eluted solution decreases with the life time of the generator due to the radioactive decay of the parent nuclide . Consequently, the useful life time of the generator is also a function of available concentration of the eluate. If we consider that the value 10–20 mCi of per mL is used as a limit of the medically useful solution, the assessment of the generator utilisation effectiveness shows the following: wasted residual activity of a used generator of 2 Ci activity eluted with 10 mL saline is 5–10% of its total activity, while smaller generators of 500 mCi activity waste up to 20–40%. In case of the concentrator used to increase the concentration of the eluate eluted from these generators, all the activity of the generator will efficiently be exploited. So, the radioisotope concentrator device should be developed to increase the concentration and quality of injectable eluates and consequently the generator life time or the effectiveness of the generator utilisation. Some concentration methods have been developed for increasing concentration of the saline eluate for extension of the life time of the fission--based generators [3844]. All these methods used a chloride-removing column containing Ag+ ions, which couple with a pertechnetate-concentrating sorbent column such as alumina, Bonelut-SAX, QMA, and multifunctional sorbent. Alternative concentration methods have also been developed. The alternatives are based on the elution of the alumina column of the generator with a nonchloride aqueous eluent (such as ammonium-acetate solution and less-chloride acetic acid solution) or with a nonchloride organic eluent (such as tributylammonium-bromide and acetone solvent). -pertechnetate of this eluate is concentrated using a sorbent column (concentration column) or an organic solvent evaporator, respectively. Then -pertechnetate is recovered in a small volume of normal saline for medical use [4560]. These methods have significantly increased the life time of the generators. The use of nonchloride eluent in replacement of saline normally used in a commercial generator may not be preferable due to legal issues of the amended registration requirement. Unfortunately, no concentrator device prototypes developed based on the developed methods are commercially available up to date. Recently, Cyclopharm Ltd. (Australia) in cooperation with Medisotec (Australia) has developed a /188Re concentrator device ULTRALUTE [4042] using a new sorbent as a concentrator column coupled with the saline-eluted commercial generator. This device (Figures 3(c) and 3(d)) is a sterile multielution cartridge which is operated/eluted by evacuated-vial through disposable sterile filters to increase the concentration of the saline eluate of aged commercial generators. The increase in concentration in the eluate enhances the utilisation of in Technegas generator-based lung perfusion (100–250 mCi/mL) and other SPECT (20–30 mCi/mL) imaging studies. The -pertechnetate of the generator eluate was concentrated more than 10-fold with a recovery yield of >85% using this radioisotope concentrator device. Five repeated elutions were successfully performed with each cartridge. So, each cartridge can be effectively used for one week in daily hospital environment for radiopharmaceutical formulation. The useful lifetime of the generator was significantly extended depending on the activity of the generator as shown in Table 2. The impurity detectable in the solution directly eluted from Gentech generator was totally eliminated by this radioisotope concentrator device and ultrapure, concentrated -pertechnetate solution was achieved. The concentrated solution is well suited to labeling in vivo kits and to loading the crucibles of Technegas aerosol generator for V/Q SPECT imaging. The useful life time of the generator (Table 2) was significantly extended from 10 to 20 days for the generators of 300–3000 mCi activity, respectively. This means that about 20% of the generator activity is saved by extending the life time of the generator. Besides that about 20% of the generator -activity can be saved as a result of the extension of -generator life time, the use of radioisotope concentrator for the optimization of generator elution to increasing the -activity yield and the effectiveness of utilization was reported by Le (2013) [58, 61]. This fact is shown as follows. continuously decays to 99Tc during his buildup from the decay of . This process not only reduces the -activity production yield of the generator (i.e. a large quantity of activity wasted during activity buildup results in a lower -activity production yield of the generator, so it is noneconomically exploited), but also makes the specific activity (SA) of continuously decreased. The low SA may cause the labelling quality of eluate degraded. This means that the elutions of the generator at a shorter build-up time of daughter nuclide will result in a higher accumulative daughter-activity production yield (more effectiveness of activity utilisation) and a better labelling quality of the generator eluate. Accumulative production yield is the sum of all the yields achieved in each early elution performed before the maximal build-up time. However, each early -elution at shorter build-up time (“early” elution) will result in a lower -elution yield and thus yields an eluate of lower -concentration because is eluted from the generator in fixed eluent volume. These facts show that a high labelling quality solution of clinically sufficient concentration could be achieved if the generator eluate obtained at an “early” elution is further concentrated by a certified radioisotope concentrator device.


Generator activity,
mCi ( GBq)
Generator useful life for SPECT imaging, daysGenerator useful life for lung imaging with Technegas, days
Without concentratorPostelution concentrator Without concentrator Postelution concentrator

100 (3.7)1601
300 (11.1)41004
500 (18.5)61206
1000 (37.0)91519
3000 (111.0)1420414

A general method described in previous work of V. S. Le and M. K. Le [58] was applied for evaluation of the effectiveness of “early” elution regime in comparison with a single elution performed at maximal build-up time point of the radionuclide generators. For this evaluation, the daughter nuclide-yield ratio is set up and calculated based on quotient of the total of daughter nuclide-elution yields () eluted in all elutions ( is the index for the elution) divided by the maximal daughter nuclide-yield or daughter nuclide-activity which could be eluted from the generator at maximal build-up time : .

Starting from the basic equation of radioactivity buildup/yield of a daughter nuclide and the maximal build-up time () for attaining the maximal activity buildup of daughter nuclide radioactivity growth-in in a given radionuclide generator system, the equation for calculation of daughter nuclide-yield ratio was derived as follows [58]:

(The subscripts and in the above equations denote the parent and daughter radionuclides, resp.).

As an example, the details of the case of generator system are briefly described as follows.

Numbers of radioactive nuclides:

Radioactivity of nuclides in the generator: the maximal build-up time (at which the maximal -activity buildup/yield in generator system is available):

Numbers of Tc atoms at build-up time: Specific activity of carrier-included in the generator system or -eluate is calculated using (3) and (5) as follows:

-Yield Ratio Calculation for Multiple “Early” Elution Regime. The value is calculated based on quotient of the total -elution yields eluted (or -activity produced/used for scans) in all elution numbers ( is the index for the th elution) divided by the maximal -activity which would be eluted from the generator at maximal build-up time .

The total -elution yields eluted in all elutions are the sum of -radioactivities at a different elution number (). This amount is described as follows: The maximal -activity buildup/yield in generator system is described using (3) and (4) as follows: -yield ratio () is derived from (7) and (8) as follows: where is the -branch decay factor of ; is the number of the early elutions needed for a practical schedule of SPECT scans. The build-up time for each elution is determined as ; is the number of the elutions which have been performed before starting a -build-up process for each consecutive elution. At this starting time point no residual Tc atoms left in the generator from a preceding elution are assumed (i.e., -elution yield of the preceding elution is assumed 100%).

The results of the evaluation (Figures 3(a) and 3(b)) based on (3), (6), and (9) show that the -activity production yield of the generator eluted with an early elution regime of build-up/elution time <6 hours increases by a factor >2 and the specific activity values of the eluates are remained higher than 160 Ci/μmol.

Obviously, the radioisotope concentrator not only may have positive impact on the extension of useful life time of the generators, but also is capable to increase both the -activity production yield of the generator/effectiveness of utilisation and the specific activity by performing the early elutions of the generator at any time before maximal buildup of .

With the utilization of concentrator device which gives a final -solution of 1.0 mL volume, the experimental results obtained from a 525 mCi generator, as an example, confirmed that the concentration and the yield of solution eluted with a 6-hour elution regime is much better than that obtained from the elution regime performed at the maximal build-up time (22.86 hours). Within first 6 days of elution, -concentration of the generator eluates is in the range 200–44 mCi/mL and total -activity eluted is 1715.7 mCi for a 6-hour elution regime (including the zero day elution) while the concentration of 83–18.2 mCi/mL and the total activity of 1015.1 mCi are for the elution regime performed at the maximal build-up time, respectively [58, 61]. The effectiveness of this early elution mode was also confirmed experimentally in the prior-of-art of generator [6264].

3. Low Specific Activity : Current Issues of Production and Prospects

generators can be produced using low specific activity . Some technologies for producing low SA have been established. Unfortunately, several alternatives are not yet commercially proven or still require further development. Presently, no nuclear reaction-based nonfission method creates a source of reasonably high or moderate specific activity. The reason is that the cross-section of all these types of nuclear reactions, which are performed by both the nuclear reactor and accelerator facility, is low ranging from several hundreds of millibarns to <11.6 barns, compared with -effective fission cross-section (37 barns) of 235U-fission used in the production of high SA as mentioned above. As shown below, SA of nonfission produced from nuclear reactor and accelerator facilities is in a range of 1–10 Ci/g Mo. To produce the generators of the same activity size (1–4 Ci) as in case of high SA mentioned above, the recovery system capable for processing Mo-target of several grams weight should be available, even though the enriched 98Mo and/or 100Mo targets are used instead of natural Mo target [2].

3.1. Production Based on Reactor Neutron Capture

Neutron capture-based production is a viable and proven technology established in the years 1960s. There are thirty-five isotopes of molybdenum known today. Of seven naturally occurring isotopes with atomic masses of 92, 94, 95, 96, 97, 98, and 100, six isotopes are stable with atomic masses from 92 to 98. 100Mo is the only naturally occurring radioactive isotope with a half-life of approximately 8.0E18 years, which decays double beta into 100Ru. All radioactive isotopes of molybdenum decay into isotopes of Nb, Tc, and Ru. 98Mo, 94Mo, and 100Mo (with natural abundance 24.1%, 9.25%, and 9.6%, resp.) are the most common isotopes used in the targetry for production of two important medical isotopes and 94Tc.

High SA cannot be produced via reaction using Mo targets because the thermal neutron cross-section for the reaction of 98Mo is relatively small at about 0.13 barn, a factor of almost 300 times less than that of the 235U fission cross-section. In this respect, irradiation of Mo targets in an epithermal neutron flux could be economically advantageous with respect to producing higher SA . The epithermal neutron capture cross-section of 98Mo is about 11.6 barn. The assessment of reaction yield and SA of the Mo targets irradiated with reactor neutrons [28, 65] shows that the irradiation time needed to reach a maximum yield and maximum SA in Mo targets is too long, while the improvement in reaction yield/SA is insignificant due to the low cross-section of 98Mo reactions. Neutron capture-based production with an 8-day irradiation in a reactor of thermal neutron flux gives a product of low SA as evaluated at EOB as follows: ~1.6 Ci /g of natural isotopic abundance molybdenum and/or 6 Ci /g of 98%-enriched 98Mo target. These values show a factor of 104 times less than that of fission-produced high SA as mentioned above. The loose-packed powder (density of > 2.5 g/cm3), pressed/sintered Mo metal powder (density of < 9.75 g/cm3), and granulated Mo metal can be used as a target material. High-density pressed/sintered 98Mo metal targets are also commercially available for the targetry. powder can be easily dissolved in sodium hydroxide. Molybdenum metallic targets can be dissolved in alkaline hydrogen peroxide or electrochemically. The metal form takes more time to dissolve than the powder form. However, the advantage of using Mo metal target is that larger weight of Mo can be irradiated in its designated irradiation position in both the research and power nuclear reactors [66, 67]. The neutron flux depression in the target may cause decreasing in production yield when a large target is used [6870]. The production capacities of 230 6-day Ci/week and 1000 6-day Ci/week are estimated for the irradiation with JMTR research reactor in Oarai and with a power reactor BWR of Hitachi-GE Nuclear Energy, Ltd., in Japan, respectively [66, 71]. The use of enriched 98Mo target material of 95% isotopic enrichment offers the product of higher SA. The W impurity in the natural Mo target material should be <10 ppm and that is not detectable in the enriched 98Mo targets. Due to high cost of highly enriched 98Mo, the economical use of this target material requires a well-established recycling of irradiated target material [2, 24, 25, 66, 67, 7274].

3.2. Accelerator Based Production

All of the accelerator-based nonfission approaches rely on highly enriched 100Mo target. While the 99% enrichment 100Mo is sufficient for all accelerator-based productions, the direct production of may require enrichments exceeding >99.5% due to the possible side reactions which generate long-lived technetium and molybdenum isotopes because these impure radionuclides would cause an unnecessary radiation dose burden to the patient and the waste disposal issues as well. The SA of   produced from the accelerators is too low for use in existing commercial generator systems that use alumina columns. New recovery technology that is suitable for processing the accelerator targets of low specific activity and allowing effective recycling of 100Mo should be developed [2].

While the specific activity of produced using accelerators (ranging up to 10 Ci/g at EOB) is not significantly higher than that of produced by neutron capture using nuclear reactor, the production using accelerator is presently focused in many research centres with regards to its safer and less costing operation compared with nuclear reactor operation. It is important to be addressed that all of the accelerator-based nonfission- production routes need a well-established technology for recycling of the 100Mo target material. This will be somewhat complicated since the 100Mo target material is contaminated with the left from the used generator systems. Handling this material presents some complicated logistics in that the target material will have to be stored until the level of is sufficiently low so as to not present radiation handling problems. Moreover, the purification of the used 100Mo target must be addressed to ensure completely removing all impurities which are brought from the chemicals and equipment used in the production processes.

3.2.1. Photon-Neutron Process 100Mo

High energy photons known as Bremsstrahlung radiation are produced by the electron beam (50 MeV electron energy with 20–100 mA current) as it interacts and loses energy in a high-Z converter target such as liquid mercury or water-cooled tungsten. The photon-neutron process is performed by directing the produced Bremsstrahlung radiation to another target material placed just behind the convertor, in this case 100Mo, to produce via the 100Mo reaction (maximal cross-section around 170 millibarns at 14.5 MeV photon energy [25]). Although the higher SA (360 Ci/g) can be achieved with a smaller weight target (~300 mg 100Mo), the produced based on a routine production base has a much lower SA, approximately 10 Ci/g [75].

3.2.2. Proton-Neutron Process 100Mo

30 MeV cyclotron can be used for production based on 100Mo reaction (maximal cross-section around 170 millibarns at 24 MeV proton energy). production yield of <50 Ci can be achieved with a bombardment current 500 mA for 24 hours [7679].

3.2.3. Neutron-Neutron Process 100Mo

production based on 100Mo reaction (maximal cross-section around 1000 millibarns at 14 MeV neutron energy) using fast neutron yielded from the reaction. The established targetry, sufficient flux of neutrons, and improvement in separation are issues to be addressed for further development [80].

3.2.4. Direct Production of

The first report on the feasibility of producing by proton irradiation of 100Mo stated that a theoretical yield of 15 Ci per hour can be achieved with 22 MeV proton bombardment at 455 μA [81]. More recently, Takács et al. found a peak cross-section of  mb at 15.7 MeV [79]. Scholten and colleagues suggested that the use of a >17 MeV cyclotron could be considered for regional production of with a production yield of 102.8 mCi/μA at saturation [78]. Estimated yield of production based on a routine production basis is 13 Ci (at EOB), using 18 MeV proton beam of 0.2 mA current for a 6-hour irradiation. A irradiation of highly enriched 100Mo target (pressed/sintered metallic 100Mo powder) using GE PET Trace cyclotron (16.5 MeV proton beam, 0.04 mA current, and 6-hour bombardment) at Cyclopet (Cyclopharm Ltd., Australia) can achieve >2.0 Ci at EOB as reported by Medisotec (Australia). Using >99.5% enriched 100Mo target produces very pure . The product of >99.6% radionuclide purity can be achieved. The major contaminants include , 95Tc, and 96Tc. Trace amounts of 95Nb are produced from the 98Mo95Nb reaction [7583].

3.3. Methods of Increasing the Specific Activity of
3.3.1. Szilard-Chalmers Recoiled

A method to increase the specific activity of neutron activated in the natural and/or enriched Mo targets using Szilard-Chalmers recoiled atom chemistry was recently reported by the scientists at the Delft University of Technology in the Netherland. The targets used in this process are 98Mo containing compounds such as molybdenum(0)hexacarbonyl [Mo(CO)6] and molybdenum(VI)dioxodioxinate [C4H3(O)–NC5H3)]2–MoO2, molybdenum nanoparticles (~100 nm), and other molybdenum tricarbonyl compounds. The neutron irradiated targets are first dissolved in an organic solvent such as dichloromethane (C2H2Cl2), chloroform (CH3Cl), benzene (C6H6), and toluene (CH3–C6H5). Then the is extracted from this target solution using an aqueous buffer solution of pH 2–12. The target material is to be recycled. This process is currently in the stage of being scaled up towards demonstration of commercial production feasibility. The specific activity of increased by a factor of more than 1000 was achieved, making the specific activity of neutron capture-based comparable to that of the high SA produced from the 235U fission. So the produced by this way can be used in existing commercial generator systems that use alumina columns [84, 85].

3.3.2. High Electric Power Off-Line Isotopic Separator for Increasing the Specific Activity of

A high power ion source coupled to a high resolution dipole magnet would be used to generate beams of Mo ions and separate the respective isotopes with the aim of producing with specific activity of greater than 1000 Ci/gram. The construction of a high power off-line isotope separator to extract high specific activity that had been produced via 98Mo and/or 100Mo routes would allow for rapid introduction of the into existing supply chain. The feedstock for the separator system will be low specific activity generated from the thermal neutron capture of 98Mo or the photon induced neutron emission on 100Mo. The proposed system would have the advantage that the produced will fit directly into the existing commercial generator system, eliminating the use of HEU and LEU targets, and can be used to generate the required target material (98Mo/100Mo) during the separation process. In addition, it can be used in conjunction with a neutron or photon sources to create a distributed low cost delivery system [2, 86].

4. Up-to-Date Technologies of Recovery from Low Specific Activity : Separation Methods, Purification/Concentration, and Generator Systems

Unfortunately, the low SA produced using the methods mentioned above contains the overwhelming excess of nonradioactive molybdenum so as the alumina columns used in existing commercial generator systems would be sufficiently loaded to produce the medically useful doses because the recovery from this source of low SA requires significantly more alumina resulting in a large elution volumes. Consequently, a solution of low -concentration is obtained from these generator systems. To make a low SA source useful for nuclear medicine application, some recovery technologies for producing medically applicable solution have been established. Unfortunately, several alternatives are not yet commercially proven or still require further development. The primary factor pertaining to the nuclear medicine scans’ quality is the concentration of in the solution produced from the generator, which is expressed as activity per mL. The injection dose activity of -based radiopharmaceuticals delivered in 1 mL solution is an important factor in determining the efficacy of the generators and the quality of -based SPECT imaging diagnosis as well. So, the recovery technologies should be developed so as a sterile injectable solution of high activity concentration and low radionuclidic and radiochemical/chemical impurity is obtained.

Up-to-date recovery technologies fall into four general categories: solvent extraction, sublimation, electrolysis, and column chromatography.

4.1. Solvent Extraction for Separation and Solvent Extraction-Based Generator Systems

Solvent extraction is the most common method for separating from low specific activity dated back to the years 1980s. The solvent extraction method can produce of high purity comparable to that obtained from alumina column-based generator loaded with fission- of high specific activity. Several extraction systems (extractant-solvent/back-extraction solution) using different extractant agents (such as ketones, crow ethers, trioctylamine, tricapryl methyl ammonium chloride (Aliquat-336), liquid ion-exchangers, and ionic liquids) were investigated [35, 60, 8791]. Among the extractant compounds investigated, methyl ethyl ketone (MEK) is the best for the extraction of -pertechnetate in terms of high extraction yield, high radiation stability, and low boiling temperature. Generators based on MEK extraction of -pertechnetate from alkaline aqueous molybdate solutions have been widely used for the production of . The extraction cycle consists of adding a mixture of MEK solvent containing 1% aqueous hydrogen peroxide to the 5 M NaOH solution of target and mechanically stirring the mixture to selectively extract the from the aqueous phase into the MEK phase. The hydrogen peroxide is added to keep the and in the appropriate oxidation state. After standing of the mixture to allow the phase separation, the supernatant MEK/ solution/organic phase containing the extracted is removed by sucking effected by a negative pressure and then it is passed through an acidic alumina to remove any that may be coextracted with into the MEK solution. In the following, the MEK/ solution is transferred to an evaporation vessel (evaporator). The evaporator is heated to ~70°C under a slight negative pressure to hasten the evaporation of the MEK. After the MEK has been completely removed, sterile saline is added to the evaporator to recover the in the form of sodium- pertechnetate dissolved in the saline. This saline solution is then sterilized by passing through a Millipore filter and transferred into a sterile vial for further processing at quality control and for formulating the radiopharmaceuticals.

The centralized solvent extraction-based generator systems have been successfully performed for more than decade in Australia [92] and Czechoslovakia [6, 35, 93, 94]. Some other systems are routinely used in Russia, Peru, and in Asian countries where the fission -based chromatographic generators do not enter the competition [60, 87, 9597]. As an example, a centralized extraction-based generator used for many years in a hospital in Vietnam is shown in Figure 4 [60].

The shortage in the fission supply today, however, has encouraged the users over the world to use more effectively the solvent extraction-based as well. So the less competitive solvent extraction-based -generator systems developed several decades before should be upgraded to be used as a user-friendly prototype for a daily use in hospital environments. The update solvent extraction-based generator systems under development are designed for an automated or semiautomated operation based either on the established extraction process [95, 98100] as mentioned above or on the improved extraction technologies. The improvement in the removing of MEK from the extracted -MEK organic phase to obtain -pertechnetate is essential in the update MEK extraction technologies, because this will make the extraction being performed with recovery into a aqueous solution without the complicated step of MEK evaporation, thus facilitating the process automation. This improved technology is based on the nonevaporation removing of MEK by passing the extracted -MEK organic phase through a cation-exchange resin or basic alumina column coupled with an acidic alumina column, followed by a water wash to completely remove both contaminant and MEK. Then the pertechnetate retained on the acidic alumina column will be eluted with a small volume of saline solution to achieve an injectable pertechnetate solution. This approach has been developed in Japan in 1971 [71, 101, 102] and recently resurrected in India and Russia [95, 99, 100]. The process is pictorially described in Figure 5. A computerized compact module for separation based MEK extraction coupled with the MEK removing unit, which composes of a tandem of basic/acidic alumina columns, is developing in BRIT [100].

4.2. Sublimation Methods for Separation and Sublimation-Based Generator Systems

Three sublimation methods for separation have been developed and commercially used in past decades [6, 35, 66, 70, 71, 92, 94, 112, 113]. The first is the high temperature sublimation method developed at the end of the sixties and used for many years in Australia, which is based on the heating a neutron-activated target on >800°C in a furnace with oxygen stream passed through. The sublimed in the form of Tc2O7 is condensed in the cold finger at the end of the furnace and is isolated by rinsing the cold finger with a hot 0.1 mM NaOH solution followed by purification on alumina. Some modified versions of this method were performed to achieve higher recovery yield. The highest yield obtained was around 80% with a sublimation time of 20–30 minutes. The second method is the medium temperature sublimation. This method relies on heating a eutectic mixture of -molybdenum oxide and metal oxides on temperature between 500 and 750°C in an air flow and ~90% of is recovered in the same way as applied in the first method. The third method is the low temperature sublimation. This method is based on the heating the solid powders of -molybdate of tetravalent metals such as titanium and zirconium molybdate on 380–450°C in a water vapour flow and 40–65% of is recovered in the saline in form of ready-to-use. Based on this method, the portable sublimation generators were commercially produced in the nineteen eighties and used for years in several hospitals in Hungary [92, 94, 114, 115]. The thermochromatographic separation at an oven temperature of 1090°C has also been successfully utilized for the recovery of from in the years 1990s [116]. This approach is expected to be used for the separation from targets. From that time until now, no update version of the sublimation-based recovery technology is found in the literature.

4.3. Electrochemical Methods for Recovery

In the past the electrochemical separation of from was performed for a radioanalytical purpose. Recently, Chakravarty et al. have further developed this method for seeking a production capability using a low specific activity . The electrodeposit and the followed pertechnetate recovery were performed at the voltage 5 V (current 500 mA and current density 300 mA/cm2) and 10 V (reversed polarity), respectively. Postelectrolysis purification of solution was also completed with an alumina column [117, 118].

4.4. Column Chromatographic Methods for Recovery and Integrated Generator Systems (Column Chromatography-Based Generator Coupled with Postelution Purification/Concentration Process)

The recovery technologies used in the separation of from low specific activity , which are based on the column chromatographic method, are recognized as the best ways to bring the low SA -based generators to the hospital users with minimal fission/nonfission Mo discrimination. Conventional chromatographic generators using alumina columns are not compatible with the loading with low SA due to its overwhelming excess of nonradioactive molybdenum. By rule of thumb, 1-2% of adsorption capacity of the alumina column loaded with molybdenum is tolerated to avoid a harmful breakthrough in the final saline eluate. To produce a generator of acceptable activity using low SA a significantly large alumina column is required to be capable to adsorb 1-2 g of Mo target, because the capacity of alumina for Mo adsorption is limited (~20 mg Mo/g of alumina). A large alumina column requires large volume of the eluent to elute patient-dose quantities of . As a consequence, large eluent volumes cause the radioactive concentration of the -pertechnetate to become unacceptably low for use in most radiopharmaceutical diagnostic procedures. So, the postelution concentration process is required to increase the -activity concentration. Although the recovery of from enriched molybdenum target material has been applied in Uzbekistan and POLATOM, the concentration of the eluate eluted from an enriched 98Mo target-based generator is moderately improved with the use of high neutron flux reactor irradiation [2].

In principle, there is no impediment for simple in-line concentration of the solution obtained from large alumina column generators using simple postelution concentration technologies. As examples, the large alumina column-based generators using low specific activity , eluted with chloride (saline) or nonchloride (acetone) eluent and combined with a concentration unit, were tested. The first low SA (7–15 GBq/g) -based generator system using up to 80-gram alumina column (jumbo alumina column generator) was developed in India [52, 53]. 70 mL saline is used for elution from this system and a concentration process with three consecutive processing steps ( loading onto Dowex-1×8 resin column; elution from the resin column with 0.2 M NaI solution; removing of ions from the effluent downstream with AgCl column) was applied. The second generator system was developed in Pakistan using a large alumina (16 g) column and acetone eluent (nonchloride organic eluent) [51]. recovery in a small volume of saline was followed after removing acetone from the /acetone eluate.

Despite the high recovery yield and good labelling quality of the highly concentrated solution achieved, the time consumption for a large volume elution and the complexity in processing at concentration stage make large alumina column-based generator systems as described above inconvincible for a commercial scale production and for the convenient utilization in the hospital environment. So, the recovery of from the low SA still requires further development to make it useful for nuclear medicine application. As a result of the development performed in many laboratories around the world, some useful recovery technologies developed up to date are described in the following.

It is the fact that the solution of high concentration cannot directly be produced from the low specific activity source, except the production based on the solvent extraction, sublimation, and electrochemical methods mentioned above. So, the technetium recovery technology based on the coupling a chromatographic -generator column of high Mo-loading capacity with a postelution purification/concentration process/unit should be considered as an important solution. This technical solution is performed by an integrated system, so-called RADIGIS (radioisotope generator integrated system) to produce a medically useful -pertechnetate solution of sufficiently high -concentration. In the following, different versions of RADIGIS developed to date are described.

4.4.1. Technetium Selective Sorbent Column-Based Recovery and Relevant Integrated Generator System

Several sorbents have been developed for selective adsorption of pertechnetate ions from aqueous solutions. Some of them, such as TEVA Spec resin (Aliquat-336 or tricapryl methyl ammonium chloride extractant impregnated in an inert substrate) and activated charcoal, adsorb ions strongly in dilute nitric acid solutions. However, the strong acidic solution (8 M HNO3) required for recovery of ions is not preferred for practical application on the basis of daily use in nuclear medicine [119123]. Some sorbents, such as ABEC (aqueous biphasic extraction chromatographic) resin and strong anion-exchange (Dowex-1×8) resin, adsorb ions from alkaline or neutral aqueous solutions. These resins are suitable for use in the production of -generator by virtue of the fact that ions can be easily desorbed from these sorbents by contacting with water or suitable organic solvent [124, 125].

(1)  Aqueous Biphasic System-Based -Pertechnetate Recovery Method [124, 126131]. A selective sorbent (ABEC-2000) column is recently developed to separate from the alkaline solution of low specific activity . A new generator system developed by NorthStar Medical Radioisotopes (USA) using low specific activity is based on the ABEC-2000 resin column coupled with an alumina guard column. This system is shown in Figure 6.

The separation process is performed as follows. An alkaline solution in 5 M NaOH obtained from dissolution of molybdenum targets is fed onto the ABEC-2000 resin column which is specifically designed to adsorb pertechnetate. Once the column is loaded, it is first washed with 5 M NaOH solution to remove any molybdate that also may have been adsorbed on the column and then by a buffer solution of pH 8. Following the wash, the technetium is stripped from the column with a normal saline solution which is then passed through an alumina guard column to remove the residual impurities. The eluate is then passed through dual 0.22 micron sterility filters to achieve an injectable -pertechnetate solution. The process can be repeated once a day as the builds up in the solution. The separation efficiencies for several consecutive days of operation were >90% with no detectable breakthrough. To date, the inherent disadvantage of this generator system reflected from the comment of user is that the elution process of this system takes a long time (about 40 minutes) and requires a 15-minute procedure for cleaning of column and tubing before the next elution is available. There is also some process to replace some components of the generator system that must be done after 5 elutions. Although the automated operation of this system facilitates the cumbersome elution-cleaning-replacing process, its being accepted as a user-friendly device may be challenged by the hospital user’s community who is quite familiar to the simple operation of the current fission -based generators.

The specific volume of solution produced by this recovery system is comparable to that of an alumina column generator loaded with the high SA fission. This new generator system is currently in the process of being validated for nuclear pharmacy use through a NDA on file with the US Food and Drug Administration [2, 130, 131].

(2)  Organic Solvent-Eluted Ion-Exchange Resin Column-Based -Pertechnetate Recovery Method. The chromatographic system of Dowex-1×8 resin column combined with tetrabutyl-ammonium-bromide (TBAB) eluent has been developed for separation of pertechnetate ions from aqueous -molybdate solution. Using commercially available anion-exchange resin Dowex-1×8 (25 mg) to selectively trap and separate from a low specific activity solution and then recovering ions from the Dowex-1×8 column by elution with TBAB in CH2Cl2 were reported. After being purified by passing through a neutral alumina column and washing the resin column with water, the alumina column will be flushed with saline to strip Na. Subsequent quality control revealed no significant levels of trace metal contaminants or organic components. recovery yields of greater than 90% were demonstrated, while radiochemical purity was consistently over 99% [125].

4.4.2. High Mo-Loading Capacity Column-Based Recovery and Relevant Integrated Generator Systems

The assessment on the capable utilisation of the high Mo-loading columns loaded with low specific activity for production of -generator is performed based on the 98Mo reaction yield and Mo-loading capacity of column packing material (). The relationship between the neutron flux of the reactor used for the production and the Mo-loading capacity () of the column packing material is derived [69, 70, 103, 132].

Based on the activation equation for the neutron capture reaction 98Mo, the activity/yield () and the relationship between and are calculated as follows:

is the Mo-loading capacity of the packing material loaded in one generator column. is the weight of molybdenum element target, which will be used for the production of one generator. is the weight of column packing material packed in one generator column. (Ci) is the given radioactivity of the generator, which is planned to be produced. is the activation time, hour. is the natural abundance of . is the molecular weight of molybdenum. hours is the halt-life of . barn is the normalised thermal and epithermal neutron activation cross-section of nuclide.

It is assumed that a generator column of the best performance for pertechnetate elution can be eluted with an eluent of volume , where () is the weight of the column packing material. The relationship between the concentration in the eluate (), the neutron flux, and is also set up. This relationship shown in Figure 7 is for a given case of the following conditions. The weight of the column packing material is 5 g and corresponding elution volume is 10 mL. The activation time of natural Mo target is l00 hours.

With these conditions, the above mentioned -equation is derived as follows:

is the Mo-loading capacity of the packing material used in the generator. is the radioactivity of in this generator. (mCi/mL) is the radioactive concentration of in the eluate eluted from the generator.

This relationship shows a general assessment on the potential use of the column packing material of given Mo-loading capacity for the -generator production using produced ex-natural molybdenum. As an example, the result assessed by above equations indicates that the column packing material of molybdenum loading capacity  mg Mo/g could be used to produce a generator of approximately 300 mCi at the generator calibration using a source of 500 mCi activity (at EOB) produced in a reactor of and thus a -pertechnetate solution of concentration <30 mCi /mL could be achieved. This solution could be used for limited numbers of organ imaging procedures due to its low concentration as shown in Table 1. With the thermal neutron flux available in the majority of the research reactors around the world, it is justified that the column packing material of  mg Mo/g should be developed for the effective use in the process of -generator production. Several sorbents, such as acidic/basic alumina, hydrous zirconium oxide, hydrous titanium oxide, manganese dioxide, silica gel, hydrotalcites, inorganic ion-exchange materials (zirconium-salt form of zirconium-phosphate ion exchanger), hydroxyapatite, mixed oxide of tetravalent metals, and diatomaceous earth, have been developed/investigated over the years [20, 133141]. These sorbents are only used for the production of fission--based -generators but they are unsuitable for -generators loaded with of low specific activity due to their low Mo-adsorption capacity (<100 mg Mo/g).

Presently, there are the limitations in the available specific activity of produced from nuclear facilities: 1–6 Ci/g Mo (1–4 Ci/g at generator calibration day) of produced in the reactors of high neutron flux using both the natural molybdenum and enriched 98Mo targets and ~10 Ci/g Mo of produced from the accelerators as mentioned above. The use of these sources and the recently developed column packing materials of high Mo-loading capacity in the process of the generator production, however, remain to be addressed. In order to reduce the solution volume eluted from a column chromatographic generator using low SA to facilitate the postelution -purification/concentration process, the columns of as high as possible Mo-loading capacity must be used. Although the Mo-loading capacity >0.25 g Mo per gram of column-packing material is achieved to date, the loading of this material with 1-2% of its capacity (similar to the loading regime of the alumina column in the fission -based generators) using a low specific available today will result in a generator of unacceptably low activity, because the produced in the majority of high neutron flux nuclear reactors and in the accelerators has a specific activity of 10000 times lower than that of the fission-based . So, the fully Mo-loaded generator columns should be used [57, 59, 60, 69, 70, 103109, 112, 113, 132, 142154]. As an example, the generated in a 4-gram weight column of high Mo-loading capacity (250 mg Mo/g), which is fully loaded with 1.0 g Mo of low specific -activity to produce a generator of 1–4 Ci on generator calibration day, can be exhaustively eluted in 10 mL saline. This eluate contains a higher breakthrough than that required for an injectable solution due to the feature of the fully Mo-loaded generator column as mentioned above. This eluate needs to be purified to remove breakthrough contaminant by passing through a sorbent column such as alumina column of ~2-gram weight. Finally, an additional volume of the saline must be used to recover all activity from the system. As a consequence, a low concentration solution of approximately 20 mL volume is produced. This value means a double of saline volume used in a fission -based generator column of 4 Ci activity loaded with 2 g alumina.

In case of the fully Mo-loaded generator columns used, the Mo affinity to the sorbent should be high enough to ensure a minimal Mo-breakthrough into the eluate eluted from the generator, because the Mo breakthrough is directly proportional with the Mo amount loaded on the column and reversely with its affinity to the sorbent (known as distribution coefficient ). To achieve a maximal affinity for the adsorption process, the chemosorption with covalent bonding between molybdate ions and functional groups of the sorbent should be expected in the process of sorbent design.

Asif and Mushtaq [155] have tested to highly load alumina column with to produce a medically acceptable pertechnetate solution of higher concentration. However, the high breakthrough in the eluate and the moderate Mo-loading capacity of this fully Mo-loaded alumina column (150 mg/g) remain inconvincible for a practical application of this technique for the generator production.

The efforts of using a fully Mo-loaded column of high Mo-loading capacity and high adsorption affinity, however, are not the all to be done in this endeavour in the process development of -generator production, because the solution volume and breakthrough of the eluate eluted from fully Mo-loaded generator columns loaded with low specific activity are still unacceptably higher compared with those obtained from the fission /alumina-based generators. All these issues suggest that the high Mo-loading capacity column-based recovery should be combined with a postelution purification/concentration process to produce a -pertechnetate solution of medically useful radioactive concentration for use in most radiopharmaceutical diagnostic procedures.

With regard to the development of generator using low SA , the column packing materials of high Mo-loading capacity developed in several laboratories are classified into two following groups. The first group includes the chemically formed solid powder materials containing molybdenum in the form of a chemical compounds such as polymolybdate compounds of tetravalent metals (in the form of solid gels) such as Zr-, Ti-, Sn-molybdates, and so forth [57, 59, 60, 69, 70, 103106, 112, 113, 132, 142147]. The second group composes of the sorbents of high Mo-adsorption capacity such as the functionalized alumina [156], the polymeric compounds of zirconium (PZC), titanium (PTC), and so forth [107, 108, 148154, 157], the nanocrystalline mixed oxides of tetravalent metals [6264, 109111, 118, 158], the nanocrystalline zirconium/titanium-oxide and alumina [159161], and recently multifunctional sorbents [4042, 58]. Such materials, as discussed below, are shown to be suitable for generator production. All these column-packing materials have a significantly higher Mo-loading capacity (>250 mg Mo per gram) than that of the alumina (‘10–20 mg Mo per gram). The can be separated from these column packings by elution with a small volume of nonsaline or saline eluents. The choice of the eluent is subject to the postelution -purification/concentration process preferred for the optimal design of an integrated system RADIGIS to produce the medically useful pertechnetate solution of sufficiently high concentration.

The chemistry of molybdate ion sorption on hydrous metal oxides is a good guide in the process of sorbent development. It is established that there are 4 adsorption sites/groups on the alumina surface: basic OH group (=Al–OH), neutral OH group (–Al–OH–Al–), acidic OH group (–Al–OH[–Al–]2), and coordinatively unsaturated site (–Al3+–). All these sites adsorb the molybdate ions to different extents depending on the pH of the solution and type of alumina sorbent used. Molybdate reacts irreversibly in a reaction (chemosorption) with the basic OH groups (at pH 8.5–6). However, as soon as these are protonated, molybdate also starts to be reversibly adsorbed by electrostatic interaction. The neutral OH groups, when protonated, also reversibly adsorb the molybdate ions. Molybdate is strongly adsorbed by the coordinatively unsaturated sites and by acidic OH groups via a physisorption/electrostatic interaction at pH <5. For this reason, acidic alumina is used for the generator production. Among tetravalent metal oxides, titania and zircona are usually used in many studies for the recovery from . Titania and possibly nanocrystalline tetragonal zircona (calcined at 600°C, IEP at the pH 4.5 [62, 156, 161]) contain mainly coordinatively unsaturated sites, so these sorbents may adsorb molybdate ions via a physisorption/electrostatic interaction at pH <5. However, hydrous titanium oxide and zirconium oxide sorbents contain many acidic and basic OH groups, respectively. Consequently molybdate ions are adsorbed on the hydrous titanium oxide surface by a physisorption mechanism at pH <4 with a less adsorption affinity compared with that of hydrous zirconium oxide which adsorbs molybdate by an irreversible chemical reaction/chemosorption. Molybdate ions adsorb on the metal oxides in different forms depending on the pH of the solution because the molybdate polymerizes in weakly acidic solution as follows:

On the polymerization, the polymerized molybdate molecules have variable molecular weights depending on the pH. This property can be experienced from the results of the potentiometric titration of molybdate solutions shown in Figure 8. As shown the molybdate is in the form of polymolybdate at pH <5 [57].

When the titanium- and zirconium-molybdate gels are used as column packing materials in the generator preparation, the molybdate covalently bonds with Ti4+ and/or Zr4+ ions in the way of nonstoichiometry. So the residual charges of the polymolybdate ions will be neutralized by the positive charge of the protons and the gels will behave as a cation exchanger. Le (1987–1994) has found the polyfunctional cation-exchange property of the titanium-and zirconium-molybdate gels [59, 69, 104]. He has taken this advantage of the molybdate gels to design the water- and organic solvent (acetone)-eluted gel-type generators as shown in Figures 14, 17, and 18 [57, 59, 60, 69, 103106, 146]. The molybdate gels have two functional groups in their structure and the total ion-exchange capacity of approximately 10 meq/g was found as shown in Figure 9. The anions, as the counter ions of the cation-exchange gel matrix, can be easily eluted with the water and water-soluble organic solvent from the column of gel-type generator. Sarkar et al. (2004) also developed a water-eluted zirconium-molybdate gel-based generator [49].

The cation exchange property can be found in all the sorbents which are fully loaded with molybdate ions. So the elution of with water or with acetone (as nonsaline eluents) from the generator column fully loaded with will provide the advantages for a consecutive postelution purification/concentration process. Le (2011) has developed an automated system of the radioisotope generator coupled with purification/concentration process using PTC/PZC sorbent columns and an eluent composed of water containing small amount of NaCl (0.005%). This system called RADIGIS- is shown in Figure 16 [6264, 109, 109111, 158].

The anions are hardly eluted from a partly Mo-loaded sorbent column with nonsaline eluents due to its strong adsorption on the unoccupied residual OH groups of the sorbent. However, this elution can be achieved if the column is wetted with a sufficient amount of residual saline. This phenomenon has been experienced in the case of the elution with acetone from an alumina column [51]. In this case the water in the aqueous saline phase existing on the sorbent surface plays a role of an ion transporter for and ions.

(1) Saline-Eluted Generator Systems Using High Mo-Loading Capacity Columns and Integrated Generator Systems

(i)  Saline-Eluted Molybdate-Gel Column-Based -Generator Systems. A zirconium-molybdate (ZrMo) and titanium-molybdate (TiMo) gels are the generator column packing materials used exclusively with low specific activity for recovery. The molybdate gel column is considered as a fully Mo-loaded sorbent column as well. These materials were first developed by Evans et al. [143] and Evans and Mattews [162] and then further improved by several research groups around the world in the 1980s [49, 57, 59, 60, 69, 70, 103106, 132, 146, 147]. A comprehensive description of molybdate gel-based generator systems using low specific activity is presented in IAEA-TECDOC-852 [70]. ZrMo and TiMo gels are prepared in the form of water insoluble solid powders containing molybdenum under a strictly controlled synthesis condition to ensure the best performance when used as a column packing material in chromatographic generators. The conditions under which a molybdate (zirconium or titanium) is prepared will influence the nanostructure of the gels and thus the generator’s performance. Different elution performances were found with the gels of amorphous or crystalline/semicrystalline structure [57, 59, 69, 132]. As a rule of thumb, the breakthrough from the generator column and the elution yield are higher with the amorphous gels, while the performance of the crystalline structure gels reverses. The porosity of the solid gel particles is also an important factor influencing the out-diffusion of the pertechnetate ions and thus the elution profile and -elution yield of the generator column. So the gel synthesis conditions such as the molar ratios of zirconium (or titanium) to molybdenum, the solution concentrations, the order of reactive agent addition, the reaction temperature, the gel aging conditions (time and temperature), the acidity of reaction mixture, the drying conditions of the gel product (time, temperature, and atmosphere), and so forth must be properly controlled in order to consistently reproduce the properties of the gel.

The