Study Finds More Than 99.9% Po-210 Removal in a Dry-Distillation Process Using the Atley C100 Module

New peer-reviewed research reports more than 99.9% removal of polonium-210 (Po-210) from irradiated bismuth target material in a dry-distillation process performed using the Atley C100 module. The result adds to the evidence available for evaluating how purification methods affect Po-210 carryover during astatine-211 (At-211) radiopharmaceutical manufacturing [1].

Why measure Po-210 removal?

At-211 is produced by irradiating bismuth-209 targets. Po-210 can arise as a by-product, both directly during irradiation and through the decay of co-produced astatine-210 (At-210). Because Po-210 is a long-lived alpha emitter, it is important to understand how much of the Po-210 present on a target could carry over into a final radiopharmaceutical dose. Isotope purification is a key step in controlling this carryover: it aims to recover At-211 from the irradiated target material while minimizing unwanted impurities such as Po-210.

Dry distillation uses heat to separate astatine from other materials in an irradiated target. Hansson and colleagues measured Po-210 carryover in a dry-distillation process performed using the Atley C100 module [1].

How the researchers measured carryover

The researchers used bismuth targets supplied by Atley Solutions and irradiated at Rigshospitalet in Copenhagen. They allowed the astatine on the targets to decay before processing them using the Atley C100 module. This enabled them to measure carryover of Po-210 already present in the target material without a contribution from At-210 that co-distilled and decayed later.

The decayed target material was processed using the Atley C100 module at 720 °C for 10 minutes at 600 mbar, with nitrogen as the carrier gas. The researchers then measured Po-210 in the material collected from the process [1].

They used two complementary analytical methods on separate targets. An alpha camera, based on a quantitative imaging technique described by Bäck and Jacobsson [4], measured Po-210 in the collected material from one target. Alpha spectrometry measured Po-210 from the other, including a subsequent rinse of the collection vial to account for material that might have remained on its walls. Because measuring Po-210 directly on these targets before processing would have interfered with the experiment, the researchers estimated their starting Po-210 content using measurements from separate irradiated targets [1].

Both methods showed very low carryover. The alpha-camera analysis found that approximately 0.087% of the estimated Po-210 on its target entered the collected material, corresponding to 99.91% separation. Alpha spectrometry found approximately 0.034% carryover from the second target, corresponding to 99.97% separation. Under the conditions tested, less than 0.1% of the estimated Po-210 present before distillation was found in the collected material [1].

How does this compare with wet chemistry?

A previous study of a wet-chemistry purification method, involving target dissolution followed by solid-phase extraction, reported more than 95% Po-210 removal [2].

The significance of purification efficiency becomes clearer in the theoretical high-activity example presented by Hansson and colleagues. Starting with an irradiated target containing 30 GBq of At-211 and an At-210 activity fraction of 0.5%, the authors calculated that approximately 309 kBq of Po-210 would form through At-210 decay during 21.6 hours, three half-lives, of transportation. Applying the conservative carryover measured in the dry-distillation study would leave approximately 268 Bq in the purified astatine fraction. Assuming 95% separation instead would leave 15.45 kBq [1].

These illustrative calculations are not a direct experimental comparison between the dry-distillation process investigated using the Atley C100 module and the wet-chemistry process. The wet-chemistry study reported more than 95% removal rather than exactly 95%.

Putting the activities in context

The IAEA lists 10 kBq as the generic exemption activity for Po-210 for moderate amounts of material [3]. In the theoretical example, 268 Bq is approximately 2.7% of that value, whereas 15.45 kBq exceeds it.

The exemption value provides context for the scale of these activities. It is not a patient-dose limit or, by itself, a measure of radiopharmaceutical safety. The paper’s theoretical example accounts for Po-210 formed through At-210 decay but does not include directly produced Po-210, for which the authors note that high-activity production data remain limited [1].

What the finding establishes

Hansson and colleagues demonstrated greater than 99.9% separation of Po-210 already present in irradiated target material for the dry-distillation process and conditions investigated using the Atley C100 module. Two analytical methods produced consistent results, with very little Po-210 found in the collected material [1].

These findings address Po-210 carryover during isotope purification. At-210 can pass through purification with At-211 and subsequently decay into Po-210. Assessing the Po-210 content of a particular final radiopharmaceutical therefore also requires consideration of its At-210 content and the time available for that decay [1].

The authors also suggest that combining dry distillation with a decentralized supply chain, in which isotope purification takes place close to the time of administration rather than before prolonged transportation of prepared doses, can help minimize Po-210 present in the dose at administration [1].

References

[1] Hansson E, Thomas R, Bäck T, Ytterbrink C, Jensen H, Lindegren S, Aneheim E. “Investigating the separation of astatine-211 and polonium-210 via dry distillation.” Journal of Radioanalytical and Nuclear Chemistry (2026). DOI: 10.1007/s10967-026-11043-3 (https://doi.org/10.1007/s10967-026-11043-3).
[2] Sevenois L, et al. “Optimised solid-phase extraction of 211At: Activity balance of 211At, 210At and 210Po after wet chemistry target dissolution.” Radiation Physics and Chemistry. 2024;225:112146.
[3] International Atomic Energy Agency. Application of the Concept of Exemption, IAEA Safety Standards Series No. GSG-17 (2023), Appendix I, Table 2. IAEA publication (https://www.iaea.org/publications/15293/application-of-the-concept-of-exemption).
[4] Bäck T, Jacobsson L. “The alpha-camera: a quantitative digital autoradiography technique using a charge-coupled device for ex vivo high-resolution bioimaging of alpha-particles.” Journal of Nuclear Medicine. 2010;51(10):1616–1623. DOI: 10.2967/jnumed.110.077578 (https://doi.org/10.2967/jnumed.110.077578).