Research ArticleSpecial Contribution

Sustained Availability of 99mTc: Possible Paths Forward

Maroor Raghavan Ambikalmajan Pillai, Ashutosh Dash and F.F. (Russ) Knapp
Journal of Nuclear Medicine February 2013, 54 (2) 313-323; DOI: https://doi.org/10.2967/jnumed.112.110338

Abstract

The availability of 99mTc for single-photon imaging in diagnostic nuclear medicine is crucial, and current availability is based on the 99Mo/99mTc generator fabricated from fission-based molybdenum (F 99Mo) produced using high enriched uranium (HEU) targets. Because of risks related to nuclear material proliferation, the use of HEU targets is being phased out and alternative strategies for production of both 99Mo and 99mTc are being evaluated intensely. There are evidently no plans for replacement of the limited number of reactors that have primarily provided most of the 99Mo. The uninterrupted, dependable availability of 99mTc is a crucial issue. For these reasons, new options being pursued include both reactor- and accelerator-based strategies to sustain the continued availability of 99mTc without the use of HEU. In this paper, the scientific and economic issues for transitioning from HEU to non-HEU are also discussed. In addition, the comparative advantages, disadvantages, technical challenges, present status, future prospects, security concerns, economic viability, and regulatory obstacles are reviewed. The international actions in progress toward evolving possible alternative strategies to produce 99Mo or 99mTc are analyzed as well. The breadth of technologies and new strategies under development to provide 99Mo and 99mTc reflects both the broad interest in and the importance of the pivotal role of 99mTc in diagnostic nuclear medicine.

Obtained from 99Mo/99mTc generators, 99mTc is the most commonly used medical radioisotope, accounting for an estimated 30 million diagnostic procedures performed annually worldwide, with approximately 50% in the United States (1–3). Roughly twenty 99mTc-labeled tracers are routinely used (4), and the demand for 99mTc is estimated to increase at an annual rate of 3%–5% (5,6). A constant and reliable supply of 99mTc is thus crucial to provide the diagnostic benefits of 99mTc-based imaging. Reactor-produced 99Mo (Fig. 1) has been the only source of 99mTc, which is mainly made by irradiation of high enriched uranium (HEU) targets. Because a transition from using HEU to low enriched uranium (LEU) is being implemented to minimize potential proliferation issues (7), other 99Mo and 99mTc production strategies must become available. Evaluation of a variety of strategies reflects the great importance of continual availability of 99mTc for nuclear medicine applications.

FIGURE 1.
FIGURE 1.

Decay scheme of 99Mo and 99mTc. T1/2 = half-life.

Table 1 summarizes information concerning the designation of uranium with respect to fissile 235U isotope content. Until 2011, more than 95% of the 99Mo required for nuclear medicine applications had been primarily produced in 7 research reactors. With the exception of the new Opal Reactor in Australia, research reactors that have been major 99Mo producers have used HEU targets (Table 2) (7,8). A few other reactors also produce fission-based molybdenum (F 99Mo) in small amounts, mainly to meet local and, at times, regional needs.

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TABLE 1

Designations for Uranium Enrichment

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TABLE 2

Operating Research Reactors Used for Large-Scale 99Mo Production

The irradiated HEU targets are processed at radiochemical laboratories that are not necessarily part of the reactor complex and in some cases are even located in different countries. The purified 99Mo solution is often shipped abroad to manufacturers for fabrication of 99Mo/99mTc generators. The most recent global disruption of 99Mo supplies began in 2007 and had wide publicity, which sensitized governments, international bodies, policy makers, physicians, patients, and the general public about the crucial role 99mTc plays in health care. Although the supply chain has stabilized for the time being, the vulnerability of depending on aging reactors for 99Mo production is expected to continue. Except for Opal, all of the reactors used for the production of 99Mo are approaching the end of their useful operational lifetimes.

Approximately 85 kg of weapon-grade HEU is used annually in targets for 99Mo production (9). Because only a small portion (∼2%–3%) of the HEU is actually involved in target fission, the current practice of not recycling targets after the removal of 99Mo means that most of the HEU remains. The proliferation risk increases over time after irradiation, because spent HEU targets can be handled more easily after the radioactive content is reduced and the shielding requirements are decreased. Minimizing the use of HEU for 99Mo production is the current target, but comprehensive abolition of the use of HEU for 99Mo production is essential to preclude potential misuse.

The U.S. Department of Energy (DOE) initiated the Reduced Enrichment for Research and Test Reactors program in 1978 to address conversion from the use of HEU as a nuclear fuel to the use of LEU (10). The Energy Policy Act of 1992 required that foreign producers of medical radioisotopes who received HEU from the United States cooperate in converting to LEU-based production (10,11) and limits U.S. export of HEU to only facilities that meet specific conditions. In 2004, the Global Threat Reduction Initiative (GTRI) was announced by U.S. Energy Secretary Spencer Abraham at the International Atomic Energy Agency headquarters in Vienna. The initiative aims to minimize as quickly as possible the amount of available weapons-usable nuclear material. The GTRI also mandates assistance for the 99Mo production facilities to help in establishing a reliable 99Mo supply network without the need for HEU (12).

The Burr amendment to the U.S. Energy Policy Act was then passed in July 2005, which permitted the export of U.S.-origin HEU to Europe and Canada for 99Mo production without any precondition for conversion to LEU (10). The 2005 Energy Act also included a provision enabling the National Academy of Sciences to ascertain “the feasibility of procuring supplies of medical isotopes from commercial sources that do not use HEU” (13). As part of its mandate, GTRI provides technical expertise, on a nonproprietary basis, to all global radioisotope producers to promote 99Mo production processes without HEU.

In January 2009, GTRI’s efforts were validated by a National Academy of Sciences report entitled “Medical Isotope Production without the Use of Highly Enriched Uranium” (14). According to this study, large-scale production of 99Mo by non-HEU methods is economically feasible. In April 2009, U.S. President Barack Obama announced a new international effort to secure all vulnerable global nuclear material within 4 y (15), which was accepted for consideration by 47 government heads of state in 2010 (16). In November 2011, the U.S. Senate passed S.99, the American Medical Isotopes Production Act of 2011 (17), intended to end U.S. reliance on foreign sources of F 99Mo. The S.99 legislation recognizes the need for stability in the supply of 99Mo for medical use and allows for domestic reactor production using LEU rather than HEU. In addition, the DOE is directed to establish a program to make LEU available, through lease contracts, for 99Mo production and retain responsibility for the final disposition of waste created by the irradiation, processing, or purification of leased uranium. Finally, this legislation makes provision to prohibit HEU exports within 7–13 y, depending on the state of the U.S. supply at that point (18).

It can be inferred that it is only a matter of time until another 99Mo crisis occurs because either U.S. HEU is curtailed to 99Mo producers or some aging reactors are shut down before alternatives are in place, or both. Although overwhelming efforts by the international community are directed toward substituting HEU with LEU for target irradiation, several other emerging options deserve serious consideration (Table 3). The production of 99Mo or 99mTc without the use of HEU will not be a trivial process and would be expected to pose formidable technical, economic, regulatory, and political challenges.

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TABLE 3

Key Strategies for 99Mo and 99mTc Production

REACTOR-BASED 99MO PRODUCTION

Fission Production of 99Mo from LEU Targets

No scientific reasons preclude the use of targets from LEU instead of HEU; however, there are technical and economic implications because of the differences in both target design and chemical processing. Conversion to LEU would reduce the 99Mo yield per mass of uranium target to approximately 20% that from HEU, but this loss is partially offset by the use of denser uranium foil targets compatible with both acidic and alkaline dissolution processes (19,20). The commercial availability of LEU foils is one of the challenges for widescale deployment of this strategy. A change in the target composition is also associated with the change in the 99Mo processing procedure, because of the higher production of unwanted 239Pu by neutron capture by 238U, which is present in 80% isotopic abundance. Although the quantity of 239Pu produced from the irradiation of LEU is still relatively small in absolute amounts, its impact on the processing of 99Mo is a point of concern. In this context, use of the Cintichem process (21,22) merits continued attention. Argentina began commercial production of 99Mo from LEU targets as early as 2002 and acquired expertise to undertake turnkey manufacturing contracts (11). With DOE assistance, South African Nuclear Energy Corporation, in South Africa, achieved the world’s first large-scale production of 99Mo in 2010 by use of LEU targets (23). Batan, in Indonesia, is also pursuing an LEU-modified Cintichem process for 99Mo production with support from Argonne National Laboratory. New LEU-based 99Mo commercial-scale production facilities have also been constructed in Egypt by Invap, S.A. (Argentina), and Pinstech in Pakistan has plans to produce 99Mo from LEU targets in a facility purchased from GSG, formerly Isotope Technologies, in Dresden, Germany (24).

Although the efforts by smaller 99Mo producers have been fruitful and have drawn widespread praise as a step in the right direction, devising an effective strategy to remove HEU from major 99Mo producers is challenging because of concerns about logistical difficulties and economics. Modifications to existing facilities require temporary shutdowns for decontamination, which in turn require that new processing facilities be available to ensure the 99Mo supply during the transition period. A key point is that any research and techniques funded by the GTRI cannot be patented by the current major 99Mo manufactures. Modifications and overhaul of the back-end processing would have to be developed independently by the manufacturers. Such options are not only expensive but also time-consuming. The fear of market share loss during significant periods of operational downtime may be another reason for the slow reaction of some 99Mo manufacturers to shift from HEU to LEU targets.

Aqueous Homogeneous Reactor Using LEU

In concept, a compact aqueous homogeneous reactor (AHR), or solution reactor, consists of 235U (LEU) in solution form as the core contained in a shielded tank or vessel. This concept precludes the need for targets; rather, at periodic intervals 99Mo can be recovered from aliquots of fuel solution removed at optimal intervals. Babcock & Wilcox Co., in association with Covidien, is currently exploring this technology for 99Mo production using a uranyl nitrate–based AHR (25). The conceptual plan consists of a 200-kW system that is estimated to yield approximately 40.7 TBq (1,100 Ci) of 99Mo per 6-d week (25). The advantages are a less complex structure, ensured inherent nuclear safety features of negative reactivity coefficients, and relatively lower costs for installation and operation. Although these positive aspects are appealing, several other equally important technologic challenges include separation process details, corrosion, uranium fuel cleanup, and waste handling. These challenges need to be addressed to ensure the technical viability of the approach. Although the AHR concept may come to fruition in the near future, this technology may be of utility in meeting long-term requirements.

Target Fuel Isotope Reactor Concept

Investigators at the U.S. Sandia National Laboratories have developed a compact 2-MW open-pool reactor fueled by LEU oxide, designated as the target-fuel isotope reactor. The proposed reactor may represent one of the lowest-risk options from the standpoints of regulatory approval and business risk and would be dedicated solely to the production of 99Mo (26). In this approach, the reactor fuel pins act as the targets for 99Mo production. The quantity of 99Mo that could be produced is directly proportional to the power attainable in the fuel pins. At periodic intervals, a chosen number of fuel pins can be withdrawn for 99Mo recovery, while simultaneously replacing with the same number of fresh fuel pins. The technology used in the concept is proven and is based on current and past research reactors and light water reactor fuel used in commercial power reactors. The safety–control system is available off the shelf from research reactor providers. The recovery of 99Mo could be accomplished with a well-known oxide dissolution process and separation procedures (27–30).

Neutron Activation Production of 99Mo

An alternative to LEU or HEU fission-based production of 99Mo is through neutron activation of molybdenum, referred to as (n,γ)99Mo production. This early established technology strategy for 99Mo production was mostly abandoned when F 99Mo became widely available (31,32). The approach provides low-specific-activity 99Mo, with specific activity ranging from 7.4 to 130 GBq/g (0.2 to 3.5 Ci/g), from reactors that have thermal neutron flux values of 5 × 1013 to 1 × 1015 neutrons·cm−2·s−1. Thermal neutrons have an energy range that allows facile interaction with target nuclei, and the neutron flux is the cross-sectional number of neutrons available for unit time. Table 4 provides a comparison of fission and neutron activation production of 99Mo. The overwhelming preference for the F 99Mo production route results from the high 235U fission cross-section (586 barn) and the 6% fission yield production of 99Mo, which together yield high 99Mo activity levels at the end of irradiation using relatively small 235U targets.

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TABLE 4

Comparison of Fission and Neutron Activation Production Routes for 99Mo

Despite these advantages for F 99Mo production, the merits of further considering the neutron activation production route are considerable. The International Atomic Energy Agency database, which provides a summary of 236 research reactors currently in operation worldwide (33), indicates that approximately 50 of these research reactors have thermal neutron flux capabilities more than 1 × 1014 n·cm−2·s−1 and that the thermal fluxes of an additional 85 reactors range from 1 × 1012 to 1 × 1014 n·cm−2·s−1. Seventy-eight of these reactors are already involved in radioisotope production, and their geographic distribution is good. Many could be used for 99Mo production by the (n,γ)99Mo route. Facilities for irradiation and postprocessing are much less technically and financially demanding than required for production of F 99Mo. The process is waste-free, and the 99Mo solution is free from fission products and actinides. Because of the better geographic distribution of such production facilities, decay loss and freight costs will be less, making the 99Mo more economical.

Use of Enriched 98Mo

The use of enriched 98Mo would be a positive step because target enrichment of 96% or greater augments the production yield and the specific activity of 99Mo by a factor of approximately 4. The low-cross-section nuclear reaction (thermal neutron cross section = 0.13 barn), however, coupled with the high cost of enriched target and the difficulties in recovering and recycling 98Mo targets, makes this route difficult to readily adapt. Maintaining a large 98Mo inventory and collecting the spent generators for recovery of enriched molybdenum targets would complicate the task, because the 99Mo/99mTc generators have a wide geographic distribution.

Power Reactor Production of 99Mo

Neutron irradiation of MoO3 in a power reactor pressure tube may also be a promising new strategy for irradiation of Mo targets. The neutron flux available in power reactors is much higher than that in research reactors. For this reason, the 99Mo produced in power reactors will have sufficient specific activity even while using natural Mo targets. Although this strategy is technically challenging compared with the use of research reactors, this option has merit because of the good distribution of power reactors throughout the world, and its potential has been assessed (34). GE Hitachi Nuclear Energy had entered into a cost-sharing cooperative agreement with the DOE’s National Nuclear Security Administration. After more than 2 y, however, GE Hitachi Nuclear Energy has recently decided not to proceed with the project, indicating that 99Mo production is not economically feasible at this time because it would compete with power generation (35). The principal challenges for widespread pursuit of (n,γ)99Mo are that the specific activity values are much lower than for fission-produced 99Mo and are dependent on the neutron flux and isotopic composition of the target used for irradiation. Furthermore, the separation methods are not as user-friendly as the column chromatography used for low-specific-activity 99Mo.

In addition to the use of traditional alumina-adsorption–type chromatographic generators, a variety of other effective methods for obtaining 99mTc from 99Mo have also been developed; these are summarized in Table 5.

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TABLE 5

Methods for Obtaining No-Carrier-Added 99mTc from Low-Specific-Activity 99Mo

Methyl Ethyl Ketone (MEK) Extraction of 99mTc from 99Mo Solution

The classic method of separating 99mTc from molybdate solution by solvent extraction using MEK is simple and capable of providing high radiochemical and radionuclidic purity and a high radioisotopic concentration of 99mTc. The MEK extraction method was abandoned in favor of alumina column generator technology when fission-produced 99Mo was available on the world market at favorable prices. This method can be quickly revived is well described (36). The concept has been implemented successfully in the Russian Federation, where it is currently used to produce 4.44 TBq (120 Ci) of 99mTc for distribution to 21 diagnostic centers in St. Petersburg (37). Using similar technology, the Medradiopreparat plant in Moscow regularly produces the 99mTc supply for various local clinics, and the Atommed Center in Moscow has developed a computer-controlled semiautomatic 99mTc delivery system based on MEK extraction of 99mTc followed by ion-exchange purification. The system, capable of handling 296 GBq (8 Ci) of 99Mo (38), includes individual small-scale units that can be operated safely at central pharmacy facilities.

Solid-Phase Column Extraction

On a similar theme, the possibility of adsorbing 99Mo on a chromatographic column containing solid-phase adsorbent followed by subsequent elution of 99mTc with MEK has also been explored (39–42). This concept, an extension of the MEK extraction procedure, is attractive as it would use (n,γ)99Mo and at the same time offer the convenience of column-based separation. This strategy thus far has been confined to laboratory-scale investigation but probably could be readily deployed for practical use.

Thermoseparation of 99mTc from 99MoO3 by Sublimation

This strategy to obtain pure 99mTc from bulky masses of (n,γ)99Mo by sublimation takes advantage of the differences in the volatilization properties of oxides of molybdenum and technetium. The option has already been investigated by some institutions (43). Such generators were in use in Australia and could produce multicurie quantities of 99mTc activity but with only 20%–25% yields. Subsequent refinement efforts in Hungary increased 99mTc yields to approximately 50% (44). Sublimation technology can be adapted for centralized production of 99mTc; however, the need to perform high-temperature operations on a regular basis with high levels of radioactivity raises safety concerns.

Zirconium Molybdate Gel Concept

Conversion of irradiated (n,γ)99Mo directly into a gel form as zirconium molybdate and loading the gel after processing into a column, followed by elution of 99mTc, is also an option (45). This widely evaluated strategy involves many intricate steps, such as dissolution, precipitation, filtration, drying, gel fragmentation, and column packing, which necessitate significant handling of radioactive material (46). The requirement for technically intense operations in a hostile radiation environment, coupled with unfavorable cost implications, has been a major pitfall for the successful use of this technology.

Preparation of Alumina-Based Generator Using (n,γ)99Mo Integrated with Postelution Concentration

Because of the limited adsorption capacity of alumina (optimal maximum, ∼20 mg of molybdenum per gram), the use of low-specific-activity 99Mo produced from 98Mo in a typical alumina-based chromatographic generator would need large-size columns to adsorb sufficient 99Mo activity. The radioactive concentration of 99mTc eluted from such columns is too low for the formulation of freeze-dried technetium kits. A postelution concentration can be used to enhance the 99Mo radioactive concentration. This strategy was originally developed to concentrate 188Re obtained from alumina-based 188W/188Re generators (47) and was successfully adapted for 99Mo/99mTc generators in the early 1990s (48).

High-Capacity Sorbent-Based Column Generator

High-capacity sorbents capable of stabilizing much larger quantities of Mo for use in column-based chromatographic systems have been developed and include a poly zirconium compound (49) and a poly titanium oxychloride (50). Synthetic alumina functionalized with a sulfate moiety (51,52) was another strategy to prepare column-based generators using (n,γ)99Mo. In recent years, nanoscale materials have caught the attention of radiopharmaceutical scientists. Because of the high surface area and intrinsic surface reactivity, nanomaterial-based sorbents possess much higher sorption capacity than conventional sorbents. Three different nanomaterial-based sorbents have been successfully exploited for the preparation of chromatographic radionuclide generators using (n,γ)99Mo (53–55). Preparation of a 13-GBq (350-mCi) generator using 99Mo with a specific activity of 14.8 GBq (400 mCi)/g was recently demonstrated (56). This strategy could be especially useful for producers having access to medium- to high-flux research reactors, wherein 99Mo of a specific activity of up to 111 GBq (3 Ci) can be obtained with natural 98Mo targets. The shielded generator assembly and the elution procedure are identical to those for existing alumina-based generators.

Electrochemical Generator

Electrochemical separation has been successful for the preparation of 90Y, 188Re, and 99mTc (57–60). Using electrochemistry for routine production of 99mTc is appealing because the electrochemical process provides separation and concentration in 1 step. This technology is adaptable even when the specific activity of 99Mo is low (3.7 GBq/g [100 mCi/g]). A fully automated system is available for the separation of clinically useful 90Y from 90Sr (61), and the technologic adaptation for making a 99Mo/99mTc generator is not expected to be difficult.

ACCELERATOR-BASED PRODUCTION OF 99MO AND 99MTC

As a major technologic alternative to the use of reactors, use of accelerators represents a promising approach for the production of 99Mo without the requirement for HEU targets. Various accelerator-based nuclear reactions that could be used to produce 99Mo (and 99mTc), along with their corresponding reaction cross-sections, are shown in Table 6. Technologic advances in the use of accelerator routes may allow the success of this concept, even though several issues related to the practical limits of irradiation volumes and the low cross-sections of the reaction routes need careful assessment. The following sections elaborate some of the accelerator options currently being considered.

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TABLE 6

Accelerator Production of 99Mo and 99mTc

Photo-Fission of Uranium Targets

Significant research effort has been expended on producing 99Mo using 238U targets to exploit the photo-fission process (62). In this process, a high-intensity beam (0.5–2 MW) of electrons is allowed to impinge on a converter target to produce bremsstrahlung photons. The photon beam is focused on the 238U target to promote fission (63). After irradiation, the uranium target is processed in the same manner as in the HEU route to recover 99Mo. According to a 2008 study by TRIUMF, in Canada, the photo-fission accelerator technique holds promise as a viable approach that has several key advantages (64). However, the extremely low cross-section of the reaction route coupled with the expense and challenges in the development of a high-power machine are major challenges for the success of this proposition (65). Nordion, also in Canada, and TRIUMF have signed an agreement to develop a technology called ZEUM (zero-enriched uranium Mo-99) to produce 99Mo by the photo-fission route. The arrangement involves state-of the-art technology to build a high-power electron accelerator, an effort that will be pursued by TRIUMF, whereas the target chemistry and analysis will be performed by Nordion (66).

Photon-Induced Transmutation by 100Mo(γ,n)99Mo Reaction

The use of high-intensity photons to initiate the 100Mo(γ,n)99Mo nuclear reaction to produce 99Mo is another strategy (Fig. 2) under consideration (65). The reaction cross-section of this path is approximately 0.1 barn at neutron energies of 12 ≤ neutron energy ≤ 17 MeV, which is much larger than with the photo-fission route and yields correspondingly increased 99Mo production. The maximum attainable specific activity of 99Mo produced through this route is estimated to reach several hundred curies of 99Mo per gram, provided very high-power accelerators are used for this purpose. TRIUMF is also evaluating the prospect of using this concept in collaboration with the National Research Council of Canada and Mevex Corp. (65). In the United States, the National Nuclear Security Administration’s GTRI has signed a cooperative agreement with North Star Medical Radioisotopes, LLC, to develop this technology to produce 99Mo and will include an operational plan, business model, and time lines (67). Although the method holds promise, some issues need to be addressed. The high-energy electron accelerators with high-power beams required for this route are currently not widely available. The enormous estimated cost for the huge inventory of enriched 100Mo that is expected to be required might be a barrier to commercialization. Substantial research and development is also necessary to develop methods to recover or recycle the 100Mo from spent generators as a means to reduce the cost of the 99Mo/99mTc generators.

FIGURE 2.
FIGURE 2.

99Mo production by 100Mo(γ,n)99Mo photon-induced reaction.

Direct Cyclotron Production of 99mTc

The direct production of 99mTc from a cyclotron beam of energetic protons using the 100Mo(p,2n)99mTc nuclear reaction may make possible the local distribution of 99mTc. This concept was initially described in 1971 (68) and has subsequently been corroborated by several researchers (69–73). Feasibility demonstration studies to produce 99mTc have been proven experimentally. It is possible to avail 2,590 GBq (70 Ci) of 99mTc in two 6-h bombardments using a high-current medium-energy (500 μA, ∼24 MeV) cyclotron (70,74,75). Direct 99mTc production using cyclotrons thus holds promise as a viable approach, even though there are issues related to the long-term availability and cost of enriched 100Mo. In addition, the requirement of a large number of cyclotrons to meet world demand and the research and development requirement associated with target design and 100Mo recycling are yet to be addressed. The coproduction of other long-lived technetium isotopes (i.e., mass dilution decreases specific activity of 99mTc) in this proposed route may require the reformulation of radiopharmaceutical kits. Although about 40 of the existing medical cyclotrons in the world could produce 99mTc by this strategy, most of these machines do not currently handle solid targets, as would be required for this route (76). The 99mTc produced by this route could meet local needs but could be distributed only regionally. Separating 99mTc from the irradiated 100Mo solid target is more complex than separating 99mTc from low-specific-activity (n,γ)99Mo. Automated modules are under development for the separation of 99mTc from the molybdenum target obtained after irradiation (77). Once the complete technology is well developed, we estimate that about 200–300 geographically well-distributed cyclotrons could satisfy world demand for 99mTc.

Accelerator-Driven Subcritical Assembly (ADS)

Another possible strategy for producing 99Mo is with an ADS, which is a unique combination of an accelerator and a subcritical nuclear reactor. The system basically consists of a proton accelerator that delivers its beam to a high-mass spallation target such as lead, tantalum, tungsten, and uranium to produce a high-intensity spallation neutron flux, which in turn is coupled to a subcritical fast core cooled with liquid metal. The spallation target is surrounded by a subcritical assembly consisting of secondary 235U (LEU) targets. Within this subcritical assembly, one can tailor the neutron spectrum of these irradiation fields for production of 99Mo. Each collision of a proton results in up to 20–30 fast neutrons (with energies mainly between 1 and 10 MeV) that in turn are moderated to produce epithermal neutrons that can be captured by 98Mo to produce 99Mo.

With this concept, it is possible to obtain (n,γ)99Mo of a relatively higher specific activity because of the reactor availability of the epithermal neutron flux profile, since the activation cross-section with epithermal neutrons is nearly 50 times higher than the thermal cross-section. In addition, 99Mo can be obtained from the chemical processing of 235U (LEU) targets of the subcritical assembly. The ADS possesses a higher level of safety because the accelerator can be turned on and off without any consequences. Joint research on this method has been undertaken by the Kharkiv Institute of Physics and Technology, in the Ukraine, and by the Belgian Nuclear Research Center. The ADS system MYRRHA in Belgium may be an alternative for nuclear reactors (78). Although the production of 99Mo using ADS may not materialize in the near future, it is of interest and utility and could eventually pave the way for a new paradigm of 99Mo production to meet future demands.

SHINE (Subcritical Hybrid Intense Neutron Emitter)

Investigators at the Morgridge Institute for Research and Phoenix Nuclear Labs, in Madison, Wisconsin, are developing a compact, relatively inexpensive system consisting of an aqueous pool of LEU nitrate or LEU sulfate solution in a subcritical assembly driven by a single deuterium-tritium beam line. Beryllium surrounding the pool provides neutron reflection and multiplication. In this system, deuterium gas is first ionized by microwaves. A low-energy (300–350 keV) DC accelerator then pushes the ions toward a target chamber containing tritium gas to produce high-energy neutrons through the D-T reaction. The neutrons then enter an aqueous LEU solution, where they multiply subcritically (Fig. 3). Uranium concentration in the solution can be controlled to maintain a subcritical pool. Unlike a reactor, this system does not create a self-sustaining nuclear chain reaction. 99Mo produced from fission of the uranium in solution can be recovered periodically (79). The advantages include lack of nuclear criticality, minimizing the chance of a criticality accident; low heat, eliminating the chance of a meltdown; and reduced nuclear waste generation. The conceptual plan of SHINE cites a 250-kW fission power system that can yield about 92,500 TBq (2,500 kCi) of 99Mo per 6-d week. The 99Mo separation in this case is identical to that from reactor-based fission routes, and a facility is expected to be completed by late 2014 (80). This option has many favorable characteristics that portend a promising future, but it has yet to reach full technologic maturity and acceptance by regulators and the user community.

FIGURE 3.
FIGURE 3.

Schematic view of ADS concept for 99Mo production.

CONCLUSIONS AND FUTURE PERSPECTIVES

With a sensible strategy, adequate resources, and sustained determination, the goal of producing 99Mo without the use of HEU targets can be achieved. Although the AHR, photo-fission, target-fuel isotope reactor, and ADS routes hold promise as innovative approaches for 99Mo production, the routine implementation of these technologies would be expected to be years away. These approaches are balanced on a fine line, with technical breakthroughs on the one hand and long-term economic viability on the other. In the medium term, accelerator production of 99Mo or 99mTc and the SHINE concept for producing fission 99Mo may hold special promise. Among the accelerator- and cyclotron-based options, direct production of 99mTc may be the most feasible and the most readily adaptable. Given the short half-life of 99mTc, however, supplying it is more akin to supplying [18F]-FDG, requiring immediate transport to the locale of use. As a result, a network of cyclotrons would need to be located proportionally to the demand, which could be synergistic with PET radionuclide production. Although most of these PET machines are in the 9- to 18-MeV range and not optimal for large-scale production of 99mTc, the existing PET cyclotrons can be upgraded to undertake large-scale production of 99mTc by use of solid targets.

Of the several non-HEU reactor options discussed, the prospect of using (n,γ)99Mo still offers appeal for its utility to address shortages of 99Mo in the immediate future. From technical and economic perspectives, the global demand for 99Mo could readily be met using (n,γ)99Mo produced in existing research reactors. These reactors would require few design changes, and they have good geographic distribution around the world. As discussed, 99mTc can be separated from (n,γ)99Mo by methods that are inexpensive, realistic, implementable in a short time frame, and capable of producing pharmaceutical-grade 99mTc.

Ensuring reliable 99Mo production without HEU is thus an evolving process in which government, manufacturers, suppliers, regulators, and users together have specific but complementary and overlapping roles and responsibilities. It is important to ensure the commercial success of non-HEU–based 99Mo production technologies to provide 99Mo or 99mTc of required quantities and quality for nuclear medicine. Otherwise, the current suppliers will move to more financially successful programs, a move that is already on the horizon as seen from their enthusiasm in developing PET-based alternatives to replace 99mTc radiopharmaceuticals (81).

Both economic strategies and regulatory aspects are intimately involved in the production of 99Mo and 99mTc. Beyond specific scientific and technical obstacles, substantial economic, political, and security issues are inhibiting the transition to non-HEU–based options for 99Mo production. The outlook for the 99Mo economy is still ill-defined in light of government subsidies and the government support of the research reactors in which target irradiations are performed. Many processing facilities that were originally funded by governments had been commercialized by the 1980s and 1990s, but they are continuing to operate with government subsidies. Over the years, even though 99Mo production technology has grown significantly and matured into a self-sustaining industry, an appropriate costing mechanism has not evolved. Such subsidies are a major hindrance, causing new entrants to doubt the economic viability of capital-intensive alternate technology to produce non-HEU–based 99Mo or 99mTc and resulting in slow HEU cleanout. All sovereign entities should relinquish the current practice of subsidies, which would alleviate the current stalemate and create new benchmarks.

Although expeditious response is needed to ensure the future availability of 99mTc, implementation of any new 99Mo or 99mTc production technologies will require approval from national regulatory agencies. Obtaining such approvals is essential, time-consuming, and expensive, but the commercial and strategic success of new production approaches depends on it. It is obligatory to ensure that any change in production strategy does not deleteriously affect radiopharmaceutical purity, efficacy, or safety.

It has been estimated that the current price of irradiating 99Mo may significantly increase in the medium term to allow new entrants to build sustainable businesses requiring investment in new non-HEU–based production facilities and to open the market to investment. An increase in the cost of 99Mo is expected to only fractionally increase the cost of 99mTc generators. The increased cost of 99mTc generators is likely to have only a marginal effect on the overall cost of diagnostic investigation because the costs of kits, scintigraphic imaging, and staff support constitute the major cost of investigations. For third-party reimbursement of typical 99mTc imaging procedures, analyses and speculations are quite clear and generally agree that any increase in the 99Mo reactor production costs, for example, would only marginally affect the total reimbursement costs for typical 99mTc imaging procedures. The 99Mo production costs represent only a small fraction of the radiopharmacy costs of 99mTc radiopharmaceutical preparations (82). Sufficient information is not yet available with regard to the accelerator production of either 99Mo or 99mTc to provide a similar analysis. For these reasons, the sustained reliable availability of 99mTc eclipses any issue regarding the total costs.

DISCLOSURE

The costs of publication of this article were defrayed in part by the payment of page charges. Therefore, and solely to indicate this fact, this article is hereby marked “advertisement” in accordance with 18 USC section 1734. Research at the Oak Ridge National Laboratory is supported by the U.S. Department of Energy under contract DE-AC05-00OR22725 with UT-Battelle, LLC. This article has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the U.S. Department of Energy. The publisher, by accepting the article for publication, acknowledges that the U.S. government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for U.S. government purposes. No other potential conflict of interest relevant to this article was reported.

Footnotes

  • Published online Dec. 19, 2012

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  • Received for publication June 21, 2012.
  • Accepted for publication September 10, 2012.
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