Sustained Availability of ^99mTc: Possible Paths Forward

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.

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.

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.

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.

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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.

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FIGURE 3.

Schematic view of ADS concept for ⁹⁹Mo production.