Diversification of ⁹⁹Mo/^99mTc Separation: Non–Fission Reactor Production of ⁹⁹Mo as a Strategy for Enhancing ^99mTc Availability

HOW MUCH ^99mTc IS REQUIRED?

Approximately 50,000 diagnostic procedures are estimated to be performed with ^99mTc radiopharmaceuticals in the United States daily (3). Patient doses vary from a few hundred megabecquerels to a maximum of 1.1 GBq. The daily activity levels of ^99mTc administered to all patients in the United States can be as high as 55.5 TBq (1,500 Ci). The total available ^99mTc activity is not used because significant activity is lost through radioactive decay, and the actual U.S. ^99mTc activity requirement could be about 111 TBq (3,000 Ci). If the proposed cyclotron route is used as the principal U.S. source of ^99mTc, it would be expected that more than 300 distributed cyclotrons may be required, whereas the use of a couple of existing nuclear reactors would be expected to suffice for our alternative strategy.

DIVERSIFIED PROCESSING FACILITIES FOR REACTOR-PRODUCED LOW-SPECIFIC-ACTIVITY ⁹⁹Mo

The reactor production of ⁹⁹Mo by irradiating natural molybdenum trioxide was used earlier for the preparation of ^99mTc but was mostly abandoned after the introduction and wide availability of the alumina-based ⁹⁹Mo/^99mTc column generator using fission-produced high-specific-activity ⁹⁹Mo. We propose to revive this earlier accepted technology using a network of geographically diversified radiochemical processing facilities for the separation of ^99mTc from ⁹⁹Mo. The irradiated molybdenum targets can be transported to these radiochemical processing facilities, and the expected transportation time is less than 24 h either by road or air. As an example, irradiation of 25 g of a ^natMo powder target (i.e., as the oxide; 24.13% ⁹⁸Mo) for 6 d in a reactor with an approximate flux of 6 × 10¹⁴ neutrons/cm²/s would result in the production of about 1.85 TBq (50 Ci) of ⁹⁹Mo at EOB. Assuming a decay loss of about 25% during transport, 1.4 TBq (38 Ci) of ⁹⁹Mo will be available for radiochemical processing. Recovering the irradiated dry powder and simple dissolution in sodium hydroxide solution provides sodium (⁹⁹Mo) molybdate, the source of ^99mTc. ^99mTc can be separated using any one of the established methods (8). Solvent extraction using methyl ethyl ketone is effective and simple, and cost-effective automated modules can be developed for the preparation of high concentrations of no-carrier-added ^99mTc activity (9). Alternatively, solid-phase extraction chromatographic separation also can be adapted (10).

Instead of a single ^99mTc extraction from the irradiated target by the cyclotron route, the ^99mTc can be extracted from ⁹⁹Mo stock solution at frequent intervals—for instance, every 6 h beginning in the early morning—and transported to users either as pertechnetate or as finished radiopharmaceuticals. The 6-h ^99mTc half-life allows easy distribution of radiopharmaceutical products. If the ⁹⁹Mo processing facilities are established in the same location as cyclotron facilities used for ¹⁸F production, the existing ¹⁸F-FDG distribution network can be used for ^99mTc products. Figure 3 shows the amount of ^99mTc available from a single batch of 1.85 TBq (50 Ci) of ⁹⁹Mo at EOB over a period of 12 d, with 3 extractions on the same day at 6-h intervals. About 9.62 TBq (260 Ci) of ^99mTc can be obtained from a single 1.85-TBq (50-Ci) batch of ⁹⁹Mo. This is in contrast to a single separation possible from a cyclotron-irradiated enriched ¹⁰⁰Mo target.

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FIGURE 3.

Decay of 1.85 TBq (50 Ci) of ⁹⁹Mo over period of 11 d and ^99mTc available on each day from single batch of 1.85 TBq (50 Ci) of ⁹⁹Mo. ^99mTc is eluted at 3 intervals starting from 400 h with gap of 6 h. Elution is not shown on day 1, since this is time required for transport of ⁹⁹Mo from reactor to processing laboratories. About 9.62 TBq (260 Ci) of ^99mTc can be recovered from single batch of 1.85 TBq (50 Ci) at EOB of ⁹⁹Mo supplied to processing center.

An important cost-saving advantage of using reactor-produced low-specific-activity ⁹⁹Mo is that usable quantities of ^99mTc activity can be repeatedly extracted, for as long as 2 wk. Fresh lots of ⁹⁹Mo could be supplied at 4-d intervals. Simultaneous operation of 3 batches of ⁹⁹Mo and multiple daily extractions could manage to produce a sustained supply of ^99mTc on a daily basis. Figure 4 illustrates the activity levels of ⁹⁹Mo available on each day and the activity of ^99mTc extractable on a daily basis. If 3 batches of ⁹⁹Mo are available simultaneously in a processing center, the ^99mTc activity available is expected to vary from 1.48 to 3.33 TBq (40–90 Ci) of ^99mTc. In such a scenario, 60 such processing facilities could produce 111 TBq (3,000 Ci) of ^99mTc, which is sufficient to conduct 50,000 daily investigations, the estimated current U.S. demand. Depending on the capacity available for irradiation, participation by a limited number of reactors in the United States could thus provide the entire U.S. ^99mTc requirement. Such a ⁹⁹Mo production scenario can be scaled up by a factor of approximately 4 with enriched ⁹⁸Mo targets (>95% ⁹⁸Mo), theoretically yielding up to 444 TBq (12,000 Ci) of ^99mTc daily. Enriched targets can be recovered for reuse, a scenario similar to the recovery and reuse of ¹⁰⁰Mo targets in the case of cyclotron production.

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FIGURE 4.

⁹⁹Mo and ^99mTc availability in central processing facility when 1.85 TBq (50 Ci) of ⁹⁹Mo are shipped at interval of 4 d. Assuming 3 elutions, ^99mTc availability at center varies from 1.48 to 3.33 TBq (40–90 Ci) per day after eighth day when 3 batches of ⁹⁹Mo are simultaneously used.

INVESTMENT NEEDED AT EACH PROCESSING FACILITY

Only moderate investment would be expected for each processing facility, with the basic essential requirements being based on the availability of licensed facilities equipped with the required shielding, ventilation, and waste-handling capabilities, which would also include 1 hot cell for processing the irradiated targets and 1 or 2 hot cells for installation of ^99mTc/⁹⁹Mo separation systems. Additional investment would be required for installation of automated modules for separation of ^99mTc from ⁹⁹Mo. The facility, as well as the hot cells used for placing the module for separation of ^99mTc, must be compliant with good manufacturing practices, since the pertechnetate will be used for the formulation of radiopharmaceuticals. Facility compliance with title 21 of Code of Federal Regulations parts 210 and 211, and U.S. Pharmacopeia <797>, may in fact be expected to be easier with the reactor route than with processing of cyclotron targets, since both methyl ethyl ketone extraction and solid-phase chromatographic procedures have already received Food and Drug Administration approval.

The expected cost for equipment to set up such a facility having 2–3 hot cells together with a couple of automated modules would be in the range of 2–4 million dollars. The production capabilities of each ^99mTc/⁹⁹Mo separation unit could be as high as 1.48–1.85 TBq (40–50 Ci) of ^99mTc per day, which translates to about 555 TBq (15,000 Ci) of ^99mTc annually, or the equivalent of 500,000 patient doses. In conjunction with existing central radiopharmacies or cyclotron centers, it is possible that the ⁹⁹Mo/^99mTc radiochemical processing units could be economically operated, since the additional operational and infrastructure costs would be minimized.

ADVANTAGES OF THE PROPOSED MODEL

Such a proposed production scenario would have several advantages that could improve the economics of the ⁹⁹Mo/^99mTc supply chain, since the concern about highly enriched uranium proliferation, waste generation, and disposal issues from the use of low enriched uranium would be avoided. The irradiation capacity of existing U.S. nuclear reactors would be used with no additional reactor facility investments, and the inherently difficult local production challenges required for direct cyclotron production of ^99mTc using enriched ¹⁰⁰Mo targets would also be avoided. Also, it remains to be seen whether the combined use of medical cyclotrons for routine production of both ¹⁸F and ^99mTc will limit the available beam time required for PET tracer development. Such an effect would be expected to reduce the ability of these facilities to develop new agents and, in turn, might negatively impact patient care. The basic radiochemical processing concepts for ⁹⁹Mo from reactor production are simpler than for processing of cyclotron targets for ^99mTc production. Although cyclotron-irradiated ¹⁰⁰Mo targets yield ^99mTc only once, reactor-irradiated ⁹⁸Mo targets provide ^99mTc available from continuous ingrowth over an extended period. Although not required for ⁹⁹Mo production from ^natMo, use of more expensive ⁹⁸Mo-enriched targets will scale up ⁹⁹Mo production by a factor of approximately 4. With enriched ⁹⁸Mo targets, the unused ⁹⁸Mo target can be recovered for future use much as is the necessary enriched ¹⁰⁰Mo recovery from cyclotron targets. There would be minimum expected loss (<25%) of ⁹⁹Mo in the proposed strategy from parent radioactive decay, since ^99mTc can be extracted from the irradiated target within 24 h after EOB, whereas decay of about 70%–80% of ⁹⁹Mo produced at EOB occurs before the initiation of alumina-based ⁹⁹Mo/^99mTc column generator use.

DISCLOSURE

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