Matyskin Artem V, Angermeier Susanna B, Drera Saleem S, Prible Michael C, Geuther Jeffrey A, Heibel Michael D
Radiation Science and Engineering Center, College of Engineering, Pennsylvania State University, 100 Breazeale Nuclear Reactor, University Park, PA 16802, United States of America.
Radiation Science and Engineering Center, College of Engineering, Pennsylvania State University, 100 Breazeale Nuclear Reactor, University Park, PA 16802, United States of America; Department of Nuclear Engineering, College of Engineering, Pennsylvania State University, 206 Hallowell Building, University Park, PA 16802, United States of America.
Nucl Med Biol. 2024 Sep-Oct;136-137:108940. doi: 10.1016/j.nucmedbio.2024.108940. Epub 2024 Jul 4.
Actinium-225 is one of the most promising radionuclides for targeted alpha therapy. Its limited availability significantly restricts clinical trials and potential applications of Ac-based radiopharmaceuticals.
In this work, we examine the possibility of Ac production from the thermal neutron flux of a nuclear reactor. For this purpose, a target consisting of 1.4 mg of Ra(NO) (T = 1600 years) and 115.5 mg of 90 % enriched, stable GdO was irradiated for 48 h in the Breazeale Nuclear Reactor with an average neutron flux of 1.7·10 cm·s. Gadolinium-157 has one of the highest thermal neutron capture cross sections of 0.25 Mb, and its neutron capture results in emission of high-energy, prompt γ-photons. Emitted γ-photons interact with Ra to produce Ra according to the Ra(γ, n)Ra reaction. Gadolinium debulking and separation of undesirable, co-produced Ac from Ra was achieved in one step by using 60 g of branched DGA resin. After Ac ingrowth from Ra (T = 14.8 d), Ac was extracted from the Ra and Ra fraction using 5 g of bDGA resin and then eluted using 5 mM HNO.
Measured activity of Ac showed that 6(1) kBq or 0.16(3) μCi (1σ) of Ra was produced at the end of bombardment from 0.9 mg of Ra.
The developed Ac separation is a waste-free process which can be used to obtain pure Ac in a nuclear reactor.
锕-225是靶向α治疗最有前景的放射性核素之一。其供应量有限严重限制了基于锕的放射性药物的临床试验和潜在应用。
在本研究中，我们研究了利用核反应堆热中子通量生产锕的可能性。为此，将一个由1.4毫克硝酸镭（半衰期T = 1600年）和115.5毫克90%富集的稳定氧化钆组成的靶，在布雷齐亚尔核反应堆中以1.7·10⁹厘米⁻²·秒⁻¹的平均中子通量辐照48小时。钆-157具有0.25兆靶恩的最高热中子俘获截面之一，其俘获中子会发射高能瞬发γ光子。发射的γ光子与镭相互作用，根据镭（γ，n）镭反应生成镭-226。通过使用60克支链二甘醇酰胺树脂一步实现了钆的减容以及从镭中分离不需要的共生锕。在镭-225从镭-226生长（半衰期T = 14.8天）后，使用5克支链二甘醇酰胺树脂从镭-226和镭-228组分中提取锕，然后用5毫摩尔/升硝酸洗脱。
测量的锕-225活度表明，在轰击结束时，从0.9毫克镭-226产生了6(1)千贝克勒尔或0.16(3)微居里（1σ）的锕-225。
所开发的锕-225分离方法是一个无废物过程，可用于在核反应堆中获得纯锕-225。
