JP4740545B2 - Method for producing a desired daughter radionuclide solution free of impurities - Google Patents

Description

(Cross-reference to related applications)
This application is a provisional application No. 60 / 372,327 filed on April 12, 2002, an application No. 10 / 159,003 filed on May 31, 2002, an application filed on September 30, 2002 Claims priority to 10 / 261,031 and application 10 / 351,717 filed on January 27, 2003.

The use of radioactive materials in diagnostic medicine is a unique structure where these procedures are safe, minimally invasive and cost effective, and the radioactive materials are not otherwise available to clinicians It is easily accepted because it provides functional and / or functional information. The use of nuclear medicine is reflected in more than 13 million diagnostic procedures performed each year in the United States alone, in other words, approximately 1 in 4 inpatients receive nuclear medical procedures. [Adelstein, at al., Eds., Isotopes for Medicine and the Life Sciences ; National Academy Press: Washington, DC, 1995; Wagner et al., “Expert Panel: Forecast Future Demand for Medical Isotopes”, Department of Energy, Office of Nuclear Energy, Science, and Technology; 1999; and Bond et al., Ind, Eng. Chem. Res. 2000, 39: 3130-3134]. More than 90% of these treatments are for diagnostic imaging purposes and use technetium- ^99m ( ^99m Tc) as the radionuclide. ^99m Tc has a unique combination of favorable productivity and availability combined with an appropriate nuclear decay mode, decay energy and chemical reactivity. These properties allow ^99m Tc to be coupled to biolocalizing agents that allow imaging of virtually all parts of many diseases and human anatomy. [Bremer, Radiochim. Acta (1987) 41: 73-81; Steigman et al., The Chemistry of Technetium in Medicine , National Academy Press: Washington, DC, (1992); Schwochau, Angew. Chem. Int. Ed. Eng (1994) 33: 2258-2267].
A typical life cycle of a pharmaceutical radionuclide such as ^99m Tc starting from raw material acquisition and leading to the nuclear synthesis of radiochemicals and clinical administration of purified and sterilized radiopharmaceuticals is shown schematically in FIG. Technetium-99m is used as a special example in this description. This is because the majority of all nuclear medicine procedures use this radionuclide and typically compare aspects of new production technologies with this successful model. ^{The 99m} Tc desired “daughter” is produced by the β ^1- (ie, negatron) decay of molybdenum-99 ( ⁹⁹ Mo) “parent” produced as a result of uranium-235 fission in the nuclear reactor. [Bremer, Radiochim. Acta (1987) 41: 73-81; Schwochau, Angew. Chem. Int. Ed. Eng. (1994) 33: 2258-2267; Boyd, Radiochim. Acta (1987) 41: 59-63; And Ali et al., Radiochim. Acta (1987) 41: 65-72].

Molybdenum-99 is separated from its nucleosynthesis precursor and by-products in “chemical processing” and represents the last step as a “radiochemical” according to FIG. Such “radiochemicals” are not subject to strict regulations on chemical and radionuclide purity and are not forced to biological conditions (eg, sterilization and non-pyrogenic). At the end of “chemical processing”, including generator production, the ⁹⁹ Mo / ^99m Tc pair becomes a “radiopharmaceutical” (according to FIG. 1), where the strictness of chemical purity, radionuclide purity, sterility and non-pyrogenicity is observed. To make adjustments.
Chemical purity is essential for safe and effective medical procedures because radionuclides are generally conjugated to biolocalizing agents prior to use. This conjugation reaction is based on the principle of coordination chemistry in which a radionuclide is chelated to a ligand that is covalently bound to a biolocalizing agent. In chemically irregular samples, the presence of ionic impurities can interfere with this conjugation reaction. If sufficient ^99m Tc is not coupled to a biolocalizing agent, for example, due to insufficient localized photon density at the target site and / or based on non-specific distribution in the blood pool or surrounding tissue Due to the rise in the body background, only poorly formed images can be obtained.
Control of radionuclide purity prevents risks associated with the introduction of long-term active or high-energy radioactive impurities into a patient, particularly when the bio-localizing agent and internal clearance properties of the radioactive impurities are unknown. Radionuclide impurities pose a great threat to patient life and such interferences are the primary focus of clinical quality control measures to prevent the administration of harmful and potentially lethal radiation doses to patients. It is.
In addition to the regulation of “radioactive” chemistry and radionuclide purity, FIG. 1 also suggests that biological conditions are established. Internal administration of a radioactive formulation clearly requires that the formulation be sterile and non-pyrogenic, and such conditions are familiar to medical practitioners.
Complementing the preferred nuclear and chemical properties of ^99m Tc is a favorable economy and convenience that allows the radionuclide to be produced to meet radiopharmaceutical specifications. In summary, these factors are essential to the success of nuclear medicine.

^The chemistry underlying the separation of ^99m Tc from ⁹⁹ Mo is the high affinity of alumina (Al ₂ O ₃ ) for molybdate-99 ( ⁹⁹ MoO ₄ ^2- ) in saline and its pertechnetate-99m ( ^99m TcO ₄ ¹⁻ ) with negligible affinity. 2, ⁹⁹ MoO ₄ ^2-parent were fixed on Al ₂ O ₃ absorbent, it from, the normally capable of conveniently separated into a vacuum container with a gradient elution of ^99m TcO ₄ ^{1-saline 99m} A Tc generator, or “ ^99m Tc cow” is shown. [Bremer, Radiochim. Acta (1987) 41: 73-81; Schwochau, Angew. Chem. Int. Ed. Eng. (1994) 33: 2258-2267; Boyd, Radiochim. Acta (1982) 30: 123-145; And Molinski, Int. J. Appl. Radiat. Isot. (1982) 33: 811-819].
The “ordinary generator” provides ^99m TcO ₄ ^1- with the appropriate chemical and radionuclide purity for use in patients, and is easy to use, compact size and solid Al ₂ O ₃ support It has the advantage of safety with the main radioactive hazard (ie ⁹⁹ MoO ₄ ²⁻ ) fixed on top. The latter advantage facilitates restrictions on the transport of generators to nuclear pharmacies and simplifies manual processing by nuclear medicine technicians.
Considering the excellent positioning of ^99m Tc in nuclear medicine and the simple and efficient operation of the conventional ^99m Tc generator shown in Figure 2, the logic and design of this radionuclide generator has become an industry standard in nuclear medicine. Yes. However, this generator methodology is not universally accepted for all radionuclides, especially those that have been proposed for use in low-specific active parent sources or therapeutic nuclear medicines. Difficulty of using the above-mentioned ordinary generator technology in low specific activity parent radionuclides, that is, trace amounts of parent radioisotopes present as a mixture with a large amount of non-radioactive parent isotopes (one or more) This comes from the need to distribute isotopes on a large volume of support so as not to exceed the absorption capacity. Large chromatographic columns are impractical in nuclear medicine applications because the desired daughter radionuclide is recovered in a large volume of eluent and therefore cannot be suitable for clinical use without secondary concentration. Radionuclides useful in therapeutic nuclear medicine present unique challenges to ordinary generator technology and deserve further discussion.

The use of radiation in disease treatment has long been put to practical use, mainly external beam radiation therapy, and is now shifting to more targeted transmission mechanisms. For example, sealed-source implants containing palladium-103 or iodine-125 have been used in prostate cancer bifurcation therapy; samarium-153 or rhenium conjugated to diphosphonate biolocalizer agent concentrates -188 has been used in the metastatic phase in osteosarcoma pain palliative therapy; radioimmunotherapy (RIT) uses radionuclide conjugates to peptides, proteins or antibodies that selectively concentrate at the disease site. Used, thereby exerting a cytotoxic effect by radioactive decay. Radioimmunotherapy represents the most selective way of delivering cytotoxic radiation doses to diseased cells while keeping healthy cells. [Whitlock, Ind. Eng. Chem. Res. (2000), 39: 3135-3139; Hassfjell et al., Chem. Rev. (2001) 101: 2019-2036; Imam, J. Radiation Oncology Biol. Phys. 2001) 51: 271-278; and McDevitt et al., Science (2001) 294: 1537-1540]. In addition, the recent surge in information about disease development and function arising from the Human Genome Project is expected to promote RIT as an advanced treatment for micrometastatic carcinomas (e.g., lymphoma and leukemia) and small to medium size tumors. Yes.
Candidate radionuclides for RIT typically have a half-life of 30 minutes to several days, coordination chemistry that allows binding to biolocalizing agents, and relatively high linear energy transfer (LET) ing. LET is defined as the energy deposited in an object per unit length of charged particles (see Choppin et al., J. Nuclear Chemistry: Theory and Applications ; Pergamon Press: Oxford, 1980). LET is substantially higher than β-particles.
For example, α-particles having an average energy in the 5-9 MeV range typically consume that energy within about 50-90 μm (corresponding to several cell diameters) in the tissue. LETβ ^1- particles with lower energy of about 0.5-2.5 MeV travel up to 10,000 μm in the tissue, and this low LET β ^1- radiation is at the cell surface to obtain a cell death probability of 99.99%. Requires as many as 100,000 collapses. However, for a single α-particle on the cell surface, its remarkably high LET gives a 20-40% probability of induction of cytotoxicity when that single α-particle passes through the cell nucleus (Hassfjell et al. ., Chem. Rev. 2001, 101: 2019-2036).

Unfortunately, LET, which makes α- and β ^1- releasing nuclides a powerful cytotoxic agent in cancer treatment, also brings many unique challenges in the production and purification of these radionuclides for use in medical applications. Yes. The most important of these challenges is the radiolysis of the support material that occurs when the conventional generator approach of FIG. 2 is used in high LET radionuclides. [Hassfjell et al., Chem. Rev. (2001) 101: 2019-2036; Gansow et al., In Radionuclide Generators: New Systems for Nuclear Medicine Applications ; Knapp et al. Eds., American Chemical Society: Washington, DC ( (1984) pp 215-227; Knapp, et al. Eds., Radionuclide Generators: New Systems for Nuclear Medicine Applications American Chemical Society: Washington, DC (1984) Vol. 241; Dietz et al., Appl. Radiat. Isot. 1992) 43: 1093-1101; Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203: 471-488; Lambrecht et al., Radiochim. Acta (1997) 77: 103-123; and Wu et al., Radiochim. Acta (1997) 79: 141-144].
Radiolysis of the generator support material comprises (a) a reduction in chemical purity (eg, radiolysis products from the support matrix can contaminate the daughter solution); (b) compromised radionuclide purity (eg, support material) Can release the parent radionuclide into the eluate (referred to as “breakthrough”)); (c) reduced daughter radionuclide yield (eg, α-recoil can cause the parent radionuclide to enter the stagnation region of the support). Pushing to make the decay product less receptive to the stripping eluate); (d) reduced column flow rate (eg, fragmentation of the support matrix increases the pressure drop towards the column) And (e) non-uniform performance (eg, product purity fluctuations, non-reproducible yields, non-uniform flow rates, etc.).

Pharmaceutical radionuclide generators typically use three basic groups of absorbents for use in the conventional manner shown in FIG. 2: (a) Organic absorbents (eg, polystyrene-divinylbenzene copolymers system ion exchange resins, polyacrylate supports for extraction chromatography), (b) inorganic absorbents (e.g., Al ₂ O _3, inorganic gels, etc.), and (c) hybrid absorber (e.g., surface grafting Inorganic frameworks containing organic chelating or ion exchange functionalities, silica supports used in extraction chromatography, etc.).
Various organic absorbents, especially conventional cation- and anion-exchange resins, have been demonstrated by their well-proven chemoselectivity [Diamond et al., In Ion Exchange , Marinsky Ed., Marcel Dekker, New York (1966) Vol. 1, p 277; and Massart, "Nuclear Science Series, Radiochemical Techniques: Cation-Exchange Techniques in Radiochemistry," NAS-NS 3113; National Academy of Sciences (1971)] and the extensive availability of these materials, Proposed for use in nuclear medicine generators [Molinski et al., Int. J. Appl. Radiat. Isot. (1982) 33: 811-819; Gansow et al., In Radionuclide Generators: New Systems for Nuclear Medicine Applications , Knapp et al. Eds., American Chemical Society, Washington, DC (1984) pp 215-227; Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203: 471-488; and Lambrecht et al., Radiochim. Acta (1997) 77: 103-123]. Unfortunately, organic ion exchange resins are frequently damaged or severely constrained in applications that use normal generator logic, typically far beyond the radiation dose required in general human applications. This is the case with very low radiation doses.
For example, polystyrene divinylbenzene copolymer-based cation exchange resins are used in generators for α-emitter ²¹² Bi, but such materials are used in 10-20 mCi generators for a “duty cycle” of approximately 2 weeks (ie, Usable generator lifetime due to chemical and physical degradation). Radiolysis of the chromatographic support results in reduced flow rates, reduced ²¹² Bi yields, and radium- ²²⁴ ( ²²⁴ Ra) parent breakthrough as reported (Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203: 471-488). Similarly, a ²¹³ Bi generator using an organic cation exchange resin is limited to a storage period of approximately 1 week at an activity value level of 2-3 mCi for the α-releasing ²²⁵ Ac parent [Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203: 471-488; and Lambrecht et al., Radiochim. Acta (1997) 77: 103-123].

According to a recent approval of the US Food and Drug Administration Yttrium -90 ⁽⁹⁰ Y) based RIT for human use extensive, efficient generator technology continues emerged from the radionuclide. Yttrium-90 is produced by β ¹ -decay of the strontium-90 ( ⁹⁰ Sr) parent radionuclide and thus represents a two-component separation involving Sr (II) and Y (III) (chemically pure) ⁹⁰ Sr product is assumed). Although various ⁹⁰ Y production methods have been proposed [Dietz et al., Appl. Radiat. Isot. (1992) 43: 1093-1101; Horwitz et al., US Pat. No. 5,368,736 (1994); and Ehrhardt et al., US Patent No. 5,154,897 (1992)], each technology has problems in scaling up to Curie level production due to problems caused by radiolysis of the solution medium and the support matrix. Inappropriateness of solvent extraction and ion exchange systems Generator for ⁹⁰ Y is as the basis for ⁹⁰ Y separation from ⁹⁰ Sr, are easily re-examined in a study has proposed a macrocyclic host / guest chemistry was . [See Dietz et al., Appl. Radiat. Isot. (1992) 43: 1093-1101; Ehrhardt et al., US Pat. No. 5,154,897 (1992)].
In these reports, ⁹⁰ Sr is separated from ⁹⁰ Y in 3M HNO ₃ on a Sr (II) selective chromatography support containing a lipophilic crown ether. This extraction chromatographic material shows unparalleled stability to gamma radiation from ⁶⁰ Co sources, but some impairment of Sr (II) retention has been observed. Unfortunately, the presence of radiolysis-induced gas pockets adversely affects the chromatographic performance of this conventional generator. Therefore, ⁹⁰ Sr is stripped after each processing operation to minimize radiolysis of the support; however, it is increasingly difficult to achieve efficient stripping of ⁹⁰ Sr during repeated use. is there.

The use of inorganic materials in radionuclide generators is highly dependent on the usual ^99m Tc generator technology based on Al ₂ O ₃ . [Bremer, Radiochim. Acta (1987) 41: 73-81; Schwochau, Angew. Chem. Int. Ed. Eng. (1994) 33: 2258-2267; Boyd, Radiochim. Acta (1987) 41: 59-63; Boyd, Radiochim. Acta (1982) 30: 123-145; Molinski, Int. J. Appl. Radiat. Isot. (1982) 33: 811-819; Benjamins et al., US Pat. No. 3,785,990 (1974); Panek-Finda US Pat. No. 3,970,583 (1976); Matthews et al., US Pat. No. 4,206,358 (1980); Benjamins et al., US Pat. No. 4,387,303 (1983); Weisner et al., US Pat. No. 4,472,299 (1984); Monze et al. ., Radiochim. Acta (1987) 41: 97-101; Forrest, US Pat. No. 4,783,305 (1988); Quint et al., US Pat. No. 4,833,329 (1989); Vanderheyden et al., US Pat. No. 4,990,787 (1991); US Pat. No. 5,109,160 (1992); Ehrhardt et al., US Pat. No. 5,382,388 (1995); and Knapp et al., US Pat. No. 5,729,821 (1998)]. While inorganic absorbers show improvements in terms of radiolytic stability, such inorganic materials often have poor ion selectivity, slow partitioning kinetics, and poorly formed morphology that inhibits good chromatographic performance. Show.
Using the ^99m Tc generator example above requires a two-component separation (ie ^99m TcO ₄ ^1- from ⁹⁹ MoO ₄ ^{2- in} saline) well adapted to Al ₂ O ₃ . However, in more complex parent-daughter relationships, several very different species may appear between the parent and daughter in a given decay chain (e.g. gas, tetravalent cation and divalent cation are ²²⁴ Ra And ²¹² Bi), it is difficult to identify a single inorganic absorbent that can hold all but the desired daughter radionuclide.

Rhenium-188 ( ¹⁸⁸ Re) is a therapeutic nuclide for the prevention of restenosis after angioplasty in osteosarcoma pain relief, as well as its coordinating chemistry with the lighter homolog Tc, which has been studied extensively Attention is being paid to certain types of RIT processing in consideration of sex. Rhenium-188 is produced by β ^1- decay of tungsten-188 ( ¹⁸⁸ W) produced by double neutron capture of enriched ¹⁸⁶ W in a high flux nuclear reactor. ^The inefficiency that occurs in ¹⁸⁸ W nucleosynthesis produces a low specific active parent; that is, trace amounts of ¹⁸⁸ W are present in large amounts of ¹⁸⁶ W isotopes. Such mass tongue state (WO ₄ ²⁻ ) requires a large column that does not exceed the capacity of Al ₂ O ₃ relative to WO ₄ ²⁻ . Large chromatographic columns provide ¹⁸⁸ Re daughters in large volumes of solution, and various secondary concentration procedures have been devised to address this shortcoming. [Knapp et al. Eds., Radionuclide Generators: New Systems for Nuclear Medicine Applications , American Chemical Society: Washington, DC (1984) Vol. 241; Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203: Lambrecht, et al., Radiochim. Acta (1997) 77: 103-123; Knapp et al., US Pat. No. 5,729,821 (1998); Knapp et al., US Pat. No. 5,186,913 (1993); and Knapp US patent No. 5,275,802 (1994)].
Another almost undiscussed disadvantage of the normal generator approach when applied to ¹⁸⁸ Re occurs after the generator has finished its duty cycle, and isotope enriched ¹⁸⁶ W is a large amount of Al ₂ O ₃ matrix. Must be extracted from. Recovery of isotope-enriched ¹⁸⁶ W in further neutron irradiation is an important part of the economical production and use of ¹⁸⁸ Re, but large amounts of isotope-enriched ¹⁸⁶ W target material on large volumes of Al ₂ O ₃ Distribution hinders cost-effective processing.

^{The 188} Re “gel generator” is trying to overcome some of the challenges facing inorganic Al ₂ O ₃ based ¹⁸⁸ Re generators, and is a highly insoluble zirconyl tungstate [ZrO (WO ₄ )] gel. Based on formation. [See Ehrhardt et al., US Pat. No. 5,382,388 (1995) and Ehrhardt et al., US Pat. No. 4,859,431 (1989)]. While this concept has several advantages over the Al ₂ O ₃ based ¹⁸⁸ Re generator, it still suffers from fundamental drawbacks when applying the usual generator approach to therapeutic radionuclides.
^{Although a 188} Re ZrO (WO ₄ ) gel generator may allow the use of a smaller column volume than an Al ₂ O ₃ based generator, the recovery of valuable isotope enriched ¹⁸⁶ W in subsequent irradiation is still complex. Additional considerations include variable chromatographic behavior and flow rates when precipitated ZrO (WO ₄ ) solids do not have a well formed particle size or morphology.
The above-mentioned inorganic materials are particularly subject to radiolysis by high LET radionuclides. Some early forms of α-releasing ²¹² Bi generators [Gansow et al., in Radionuclide Generators: New Systems for Nuclear Medicine Applications ; Knapp et al. Eds., American Chemical Society: Washington, DC (1984) pp 215- 227; and Mirzadeh, S. Generator-Produced Alpha-Emitters. Appl. Radiat. Isot. (1998) 49: 345-349] uses inorganic titanates to retain the long-lived thorium-228 parent, From that, ²²⁴ Ra daughter was eluted and then adsorbed on a normal cation exchange resin. Over time, the titanate column material yielded fine particulates that yielded radiolysis and were forced to separate under pressure.
The hybrid absorbent can be subdivided into an extraction chromatography material and a processed inorganic ion exchange material. In most published applications of hybrid materials, well-known extraction chromatography methods are used [Dietz et al., In Metal Ion Separation and Preconcentration: Progress and Opportunities ; Bond et al. Eds., American Chemical Society, Washington, DC (1999) Vol. 716, pp 234-250], while the production and use of processed inorganic materials is a fairly recent phenomenon. Extraction chromatography overcomes the poor ion selectivity and slow separation kinetics of inorganic materials by using solvent extraction reagents physisorbed onto an inert chromatography substrate. [See Dietz et al., In Metal Ion Separation and Preconcentration: Progress and Opportunities ; Bond et al. Eds., American Chemical Society, Washington, DC (1999) Vol. 716, pp 234-250].

The radiolytic stability of an extraction chromatography support is the most significant result reflected as a sustained flow rate over the generator duty cycle when its inert substrate is an amorphous inorganic material such as silica. Improved by. However, such “improved” radiolytic stability is still associated with molecules constructed from organic frameworks where the basic chemical reactions underlying parent / daughter separation remain sensitive to radiolysis. It ’s just an appearance. Similarly, organic chelating components have also been introduced into processed inorganic ion exchange materials to improve ion selectivity, but such functionality continues to be subject to radiolysis.
Preliminary reports have been presented using hybrid absorbents as conventional generator supports in the production of ²¹³ Bi. [See Lambrecht et al., Radiochim. Acta (1997) 77: 103-123; Wu et al., Radiochim. Acta (1997) 79: 141-144; and Horwitz et al., US Pat. No. 5,854,968 (1998)]. Early work relies on adsorption of ²²⁵ Ra by an organic cation exchange resin, which shows substantial degradation in a short time, with reduced yield of ²¹³ Bi, poor radionuclide purity and unacceptably slow column Only the flow rate was obtained. [See Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203: 471-488; and Lambrecht, et al., Radiochim. Acta (1997) 77: 103-123]. Early improvements centered on Dipex ^R Resin, an adsorption of ²¹³ Bi ²²⁵ Ac on an inert silica gel support that was physically adsorbed with diphosphonic diester for chelation. [See Horwitz et al., React. Funct. Polymers (1997) 33: 25-36]. This silica substrate exhibits higher radiolytic stability than previously used organic cation exchange resins; however, radiolytic degradation (ie, discoloration) is around a narrow chromatographic band loading the ²²⁵ Ac parent. It was observed at the end, and eventually it reached a breakthrough of ²²⁵ Ac parents. [See Lambrecht et al., Radiochim. Acta (1997) 77: 103-123; and Wu et al., Radiochim. Acta (1997) 79: 141-144].
Additional improvements in the generator include ²²⁵ Ac parent radiation over a larger volume of chromatography support achieved by batch loading Dipex ^R Resin with ²²⁵ Ac rather than in a narrow chromatography band. The focus was on reducing the radiation density by dispersing the activity. [See Wu et al., Radiochim. Acta (1997) 79: 141-144]. Unfortunately, this batch loading method is cumbersome and Dipex ^R Resin is still subject to radiolysis of the chelating diphosphonic diester, on which the separation efficiency depends.

Despite the industrial preemptive nature of the conventional generator shown in FIG. 2, the basic limitation described above is the radiolysis of the support medium when using high levels of high LET radioactivity useful in therapeutic nuclear medicine. Has been exacerbated by. The importance of these constraints combined with the ultimate responsibility of compromised patient safety urges the development of alternative generator technologies for radionuclides that are therapeutically useful, among others.
The ideal generator technology should provide reliable production close to the theoretical yield of the desired daughter radionuclide with operational simplicity and convenience and high chemical and radionuclide purity. As developed in diagnostic radionuclides, conventional generator technology generally meets some of these criteria, but it has been observed that purity and yield vary. [See Boyd, Radiochim. Acta (1982) 30: 123-145; and Molinski, Int. J. Appl. Radiat. Isot. (1982) 33: 811-819].
However, normal generators are only insufficient for systems containing low specific activity parents (e.g., the ¹⁸⁸ W / ¹⁸⁸ Re generator described above) as well as systems containing high LET radionuclides useful in therapeutic nuclear medicine. Not suitable. In order to safely and reliably produce therapeutically useful radionuclides with high chemical and radionuclide purity, a new paradigm in radionuclide generator technology is required. A shift in the basic principles of managing generator technology for nuclear medicine, especially therapeutic nuclides, is the potential for careless administration of long-lived parents of high-LET therapeutic radionuclides, potentially compromising the already weak health of patients. It is further supported by the fact that it leads to death. The conventional generator approach shown in FIG. 2 relies on long-term storage of the parent radionuclide on a solid support that is routinely subjected to high LET radiation, so that the daughter over an approximately 14-60 day generator duty cycle. No guarantees regarding radionuclide chemistry or radionuclide purity can be made.

  Further support for fundamental changes in radionuclide generator technology comes from the rapidly growing trend towards automation of common tasks such as synthesis operations in biotechnology as well as from high-throughput blood screening in clinical laboratories. Has occurred. Radionuclide generator technology, such as that implemented at nuclear pharmacies, currently lags behind in automating daily activities. In the field of nuclear medicine, the industry appears to be moving in the direction of automation due to government regulations as a measure to protect patient health and the competition / profitability of companies. The introduction of a computer controlled liquid delivery system at each pharmacy would allow a departure from the vacuum vessel system generator of FIG. Many manual reductions also work to minimize radiation doses to nuclear medicine technicians, while at the same time reducing liability due to human error.

  The harmful effects of radiolysis described above pose significant challenges in the development of new therapeutic radionuclide generators. Any damage to the normal generator support material compromises the separation efficiency, potentially causing a breakthrough of the parent radionuclide, leading to a potentially lethal dose of radiation when administered to the patient. Although such catastrophic events are theoretically prevented by quality control measures integrated into nuclear drug product operations, any lack of safe and predictable generator behavior can be attributed to nuclear pharmacies, hospitals and their It represents a great responsibility for funders. The present invention described below provides an alternative radionuclide generator technology that can reliably produce medically useful radionuclides with high chemical and radionuclide purity close to the theoretical yield.

The present invention contemplates a method for producing a desired daughter radionuclide solution that is substantially free of impurities. The method comprises an aqueous parent-daughter radionuclide containing a desired daughter radionuclide, a first separation medium having a high affinity for the desired daughter radionuclide and a low affinity for the parent and other daughter radionuclides; A step of contacting. The parent and desired daughter radionuclide have different ionic charges and / or different charge densities when present in the solution. The contact is maintained for a period of time sufficient for the desired daughter radionuclide to bind to the first separation medium to provide a solution containing the desired daughter-carrying separation medium and a reduced concentration of the desired daughter radionuclide (initial parent-daughter radionuclide solution). Compared to).
The solution containing a reduced concentration of the desired daughter radionuclide is removed from the desired daughter-carrying separation medium. The desired daughter radionuclide is stripped from the desired daughter-carrying separation medium to produce a desired daughter radionuclide solution. The desired daughter radionuclide solution is contacted with a second separation medium having a high affinity for the parent radionuclide and a low affinity for the desired daughter radionuclide. In a preferred embodiment, no chemical conditioning is performed on the solution prior to elution on the second separation medium (Guard column). The contact is maintained for a time sufficient for the parent radionuclide, if present, to bind to the second separation medium to produce a desired daughter radionuclide solution that is substantially free of impurities. The daughter radionuclide solution that is substantially free of impurities is typically recovered, but the solution can be used without recovery in reactions such as binding of the radionuclide with a medically useful agent.

The present invention has several benefits and advantages.
In one benefit, the above method does not require the use of air or gas to separate parts of the solution from each other, in other words, better chromatographic operating performance and better overall chemistry. And providing radionuclide purity.
The advantage of the intended method is that the separation medium has a longer usable life since it has a tendency to not be decomposed by radiation due to the relatively short time it takes to contact the high-energy energized radionuclide and the separation medium. It is to have.
Another advantage of the present invention is that radionuclides with high purity can be obtained.
Another advantage of the present invention is that wider tolerances can be utilized in the selection of commercially available separation media pairs, and appropriate elution solutions can be used for the production of various radionuclides for medical and analytical applications. It can be easily prepared.
A further advantage of the present invention is that the daughter radionuclide can be recovered with a small volume of eluent solution due to the high separation efficiency of the separation medium.
A further advantage of the present invention is that the chemical integrity of the separation medium is retained, which provides a more predictable separation performance and reduces the potential for parental radionuclide contamination of the daughter product. .
Further benefits and advantages will be readily apparent to those skilled in the art from the following description.

Answers to the problems posed by radiolysis when using high LET radionuclides use a method generally referred to herein as multi-column selectivity reversal from a solution containing both parent and desired daughter radionuclides. To be separated in the present invention. The term “parent radionuclide” refers to one or more parent radionuclides and a desired daughter nuclide and its daughter nuclide whose intended solution containing the parent and desired daughter radionuclide is well known from the radioactive decay series. For the purposes of understanding that daughter nuclides may or may not be included, they are often used herein in the singular.
The contemplated method preferably uses multiple chromatography columns in the above separation. The separation media packing of these columns has different selectivities for the parent and desired daughter radionuclides, and these selectivities are commonly used in similar separations as performed in the normal generator approach of FIG. It reverses from selectivity. That is, the first separation medium that is contacted with the aqueous solution containing the parent and the desired daughter has a greater selectivity for the desired daughter than for the parent or other daughters that may be present, while at least once later. The separation medium to be contacted has a greater selectivity for the parent than for the desired daughter. It should be noted that multiple second separation media can be used in a single separation, and these media can be separate or identical guard columns suitable for the particular medium used.
Radioactive parent and daughter solution storage reduces the radiolysis of chromatographic separation materials that contribute to product purity because the majority of radiolytic degradation is transferred to the solution matrix, e.g. water, rather than to the separation medium. Has a significant advantage of minimizing.

The integrity of the separation medium is further maintained by minimizing contact time between the radioactive solution and the selective separation medium for the daughter radionuclide using a high chromatographic flow rate (e.g., via an automated fluid delivery system). Is done. Maintaining the chemical integrity of the separation medium is equated with a more predictable separation performance and reduces the likelihood that the parent radionuclide will contaminate the daughter product. Furthermore, by targeting the extraction of the desired daughter radionuclide as needed rather than by elution, no conventional generators or inorganic absorbers resistant to radiolysis are required, with greater solute selectivity More types of chromatographic separation media can be used.
In order to further minimize the possibility of parent radionuclide contamination, another separation medium selective to the parent (one or more) is introduced downstream of the desired daughter selective separation medium. The addition of a second separation column adds further safety and ensures that dangerous long-lived parent radionuclides are not administered to the patient. An example of such a serial column arrangement is shown in FIG. Examples of desired daughter ion / parent ion groups that can be easily separated using the method of the present invention include Y ³⁺ / Sr ²⁺ ; TcO ₄ ¹⁻ / MoO ₄ ²⁻ ; PdCl ₄ ²⁻ / Rh ³⁺ In ³⁺ / Cd ²⁺ ; I ¹⁻ / Sb ³⁺ ; ReO ₄ ¹⁻ / WO ₄ ²⁻ ; Tl ¹⁺ / Pb ²⁺ ; Sc ³⁺ / TiO ²⁺ or Ti ⁴⁺ ; Bi ³⁺ / Ra ²⁺ , Pb ²⁺ ; Bi ³⁺ / Ac ³⁺ , Ra ²⁺ ; At ¹⁻ / Bi ³⁺ ; and Ra ²⁺ / Ac ³⁺ , Th ⁴⁺ .
As shown at the top of FIG. 3, the parent and desired daughter radionuclides approach or reach a radioactive steady state in an aqueous matrix that is irradiated with a radiation dose rather than on a separation medium that contributes to the efficiency of chemical separation. Let me. When necessary, the solution containing the parent and desired daughter radionuclide is contacted with a chromatography column (primary separation column) containing a first separation medium selective for the daughter radionuclide (on the column). While eluting any other “daughter” such as one or more parents and the daughter of the desired daughter radionuclide. The desired daughter and the one or more parent radionuclides have one or both of different (i) ionic charges or (ii) charge densities when present in the solution.
That is, with respect to ionic charge, when present in the solution used to contact the first separation medium, one of the parent and daughter radionuclides can be +2 cations and the other can be +3 cations, or one can be A +2 cation and the other can be a -1 anion, etc. Typically, the parent and desired daughter radionuclide maintain, but need not maintain, charge differences throughout the separation process. For example, if TcO ₄ ¹⁻ is separated from MoO ₄ ²⁻ or ReO ₄ ¹⁻ is separated from WO ₄ ²⁻ , these anions retain their charge throughout the separation. On the other hand, both bismuth and actinium typically have a +3 charge, whereas bismuth is preferentially separated from actinium as a solution complex with a chloride ion such as BiCl ₄ ^1- anion, whereas actinium. Does not form such a complex under the same conditions and remains as an Ac ³⁺ cation.

Although most chemical separations can be conveniently explained by recognizing the difference in net ionic charge of two or more analytes as the basis for the separation, many other separations can be coordinated chemistry as a means of performing the separation. And / or rely on more subtle differences in solution seed formation. As a general approximation, the difference in coordination preference and / or solution species formation between two ions can be conveniently attributed to different charge densities where the electrostatic interaction is dominant.
Charge density is defined as the overall charge per unit volume occupied by a given mono- or polyatomic ion. The concept of charge density is a contributing factor to the hard / soft acid / base theory. According to this theory, ions defined as “hard” are not very polarizable and typically have large charge density absolute values (eg, Li ⁺ , Al ³⁺ , F ⁻ , and O ^{2). -),} whereas ions defined as "soft" has a lower charge density are easily polarized ^{(e.g., Hg 2+, Bi 3+, I} 1-, TcO 4 1- etc).
An explanation based solely on the difference in ionic charge is the separation of many similarly charged analytes that are typically separated based on differences in analyte charge density, for example, separation of Ce ³⁺ from Lu ³⁺ Or the separation of F ^1- from I ^1- is not adequately explained. In Ce ³⁺ / Lu ³⁺ separations, each anion has the same charge, but the well-known lanthanide contraction results in a systematic decrease in lanthanide ion radius and thus ion capacity, resulting in a net increase in charge density across the lanthanide series. Is produced. This net increase in charge density can lead to differences in hydration numbers (primary and secondary ranges), solution species formation and coordination chemistry that can act individually or collectively as a basis for separation.
In another example, the charge density of the halide anion decreases as the group transitions as the ionic radius (and capacity) increases and the charge becomes more diffuse. Such charge density differences can be exploited in separations because the electrostatic interactions that govern ion-ligand and ion-solvent interactions are different and provide a convenient chemical aspect to be utilized in a given separation.
The concept of charge density is not strictly limited to monoatomic ions and easily extends to polyatomic species; for example, NH ₄ ¹⁺ / N (CH ₂ CH ₃ ) ¹⁺ and TcO ₄ ¹⁻ / IO ₃ ^1- . In each instance, each ion has a similar charge, but each occupies a different capacity, thereby changing the charge density, such as hydration free energy, overall hydration number, complex formation constant, etc. Alter ionic interaction properties and solution species formation as reflected in the parameters.

The eluate from the primary separation column containing the parent and a reduced amount of the desired daughter radionuclide (the desired daughter-deficient parent-daughter solution or the solution containing the reduced concentration of the desired daughter radionuclide) is the first separation carrying the desired daughter. Remove (separate) from the medium. This solution can be discarded, but preferably it can be collected in a container and again brought close to a radioactive steady state to obtain additional quantities of the desired daughter. The primary separation column containing the daughter radionuclide is then typically washed to remove any residual impurities that may be present, such as from gaps, prior to daughter elution (stripping).
To maximize the convenience and effectiveness of this multi-column generator approach, the knowledge of the daughter radionuclide and the solution species formation of its radionuclide parent can be used to determine the strip solution and the second chromatography column (guard column). Select both materials (one or more) of the second separation medium. In an ideal implementation, the daughter-selective primary separation medium-containing column allows the desired daughter radionuclide to elute directly into the guard column without any chemical adjustment to the solution medium. While the parent or other daughter ion interferers that may be present are retained on the second column.
The solution storage of the radioactive raw material, as well as the use of a multi-column selective inversion method in which the desired daughter radionuclide is first selectively extracted and then the residual parent ions are further decontaminated by a guard column containing a second separation medium, Enabling the reliable production of high purity desired daughter radionuclides that function to minimize sex degradation and are close to the theoretical yield. In a typical application, the primary separation column exhibits high affinity for the desired daughter and low affinity for the parent and any other daughter radionuclide, while the guard column has high affinity for the parent and the desired daughter. Contains a second separation medium having low affinity for the radionuclide.

放射性定常状態の系(例えば、²²⁴Ra及びそれらの娘(²¹²Bi及びその娘を含む)においては、その分母は約1である。このことは、DF値を、クロマトグラフ中のストリッピングピークを調べ分析物(即ち、所望²¹²Bi娘放射性核種)における最大cpm/mLを不純物(即ち、²²⁴Ra親)の活量で割ることによって概算し得ることを意味する。
また、DFは、分析物と不純物における乾燥重量分布比(D_w)の比率を取ることによっても算出し得る。“流入液”が放射性定常状態にある(DFユニティーの分母をなす)と仮定すれば、分析物/不純物におけるD_w値の比は、下記のとおりであり、

これを約分して、下記に簡単化する：

(式中、A_o、A_f、V、m_R及び％固形分は、別途定義したとおりである)。
これらの活量比は、DFの定義において別途述べたモル濃度に比例している。 Such a combination provides a parent total decontamination factor (DF) from about 10 ⁴ to about 10 ¹⁰ or more desired daughter radionuclides under conditions of contact with the multiple separation media. Separately, each column used exhibits about 10 ² to about 10 ⁵ or more DF under the contact conditions. When the DF in one predetermined process is multiplied by the DF in the next process or expressed using an index, the DF value index is added in each process. A DF value of about 10 ¹⁰ is approximately the maximum DF that can be easily measured using typical radioactivity analysis laboratory equipment.
Decontamination factor (DF) is defined using the following equation:

In radioactive steady-state systems (eg, ²²⁴ Ra and their daughters (including ²¹² Bi and its daughters)), the denominator is about 1. This means that the DF value is determined by the stripping peak in the chromatograph. It means that it can be estimated by dividing the maximum cpm / mL in the probe analyte (ie the desired ²¹² Bi daughter radionuclide) by the activity of the impurity (ie ²²⁴ Ra parent).
DF can also be calculated by taking the ratio of dry weight distribution ratio (D _w ) between analyte and impurities. Assuming that the “influent” is in a radioactive steady state (which forms the denominator of DF unity), the ratio of D _w values in the analyte / impurity is:

Reduce this and simplify it to:

( _Where A _o , A _f , V, m _R and% solids are as defined separately).
These activity ratios are proportional to the molar concentrations mentioned separately in the definition of DF.

Thus, there are at least three basic differences between the intended multi-column selective inversion generator technique and the conventional approach shown in FIG. 2: (1) The storage medium for the parent radionuclide is more solid than the solid support. (2) The desired daughter radionuclide is selectively extracted from the parent radionuclide containing solution when needed; (3) the second separation medium is from the generator system of the parent radionuclide. Prevent release.
In addition to minimizing radiolytic degradation to the chromatographic support, the extraction of micromass daughters (ie, microcomponents) using the multi-column selective inversion generator shown in FIG. Enable. That is, the desired daughter radionuclide can be recovered in a small volume of solution conveniently diluted to a dose suitable for clinical use. Typically, 90% of the daughter radionuclide can be transmitted in less than about 5 bed volumes of the first separation medium in the first column.
The intended separation method is typically carried out at ambient room temperature. Gravity flow to each column can be used, but preferably the separation is carried out at 1 atm or higher as obtained by a hand-operated syringe or power pump. Also preferred is the use of less than 1 atmosphere achieved by the use of a syringe (eg, vacuum assisted flow).
The contact time between the solution and the separation medium is typically the residence time of the solution through the column, regardless of the use of a pressure head. That is, a given solution and separation medium can be mixed and the resulting contact can be maintained for hours or days, but adsorption by the separation medium is usually sufficiently rapid; that is, binding and phase transfer reactions. Is sufficiently rapid and the contact obtained by the flow on and in the separation medium particles provides sufficient contact time to effect the desired separation.

The general concept of selectivity reversal between extraction of the desired daughter radionuclide with a primary separation column and retention of parent and other interferents with a guard column represents an important aspect of the present invention. An apparently similar concept has been proposed for use with diagnostic ⁶⁴ Cu radionuclides [see Zinn, US Pat. No. 5,409,677 (1995)]; however, for the radiotherapeutic radionuclides of multi-column selective reversal generators. Alternatively, applications for high specific activity diagnostic radionuclides have not been previously tested or recognized, and the ionic charges of both parent and daughter radionuclides in the above disclosure are the same +2 for copper and zinc ions. The charge densities of Cu ²⁺ and Zn ²⁺ ions are substantially the same.
That is, the examples cited for ⁶⁴ Cu relies exclusively on the use of immobilized ligands to complex ⁶⁴ Cu, are removed ⁶⁴ Cu from a large amount of zinc isotopes. One mention of secondary removal of zinc by an unidentified anion exchange resin from ⁶⁴ Cu products is made necessary by the poor selectivity exhibited by complexed ligands in the initial separation. Furthermore, a large bed volume is required, and the ⁶⁴ Cu product is transferred into> 20 mL of strongly acidic solution, which can conjugate ⁶⁴ Cu to a biolocalizer for use in medical procedures. Requires secondary concentration and neutralization before it becomes. The proposed ⁶⁴ Cu separation system also provides high specific activity radionuclide generators or high LET radiation (both present unique challenges to the design of radionuclide generators) for the identity of each ion to be separated ) Does not explain any application in use.

If there is little concern about the radiolysis of the support material (eg, in diagnostic radionuclides), the multi-column selective inversion generator shown in FIG. 3 continues to provide many advantages. For example, target irradiation in an accelerator or reactor often requires the use of isotope enriched target material to maximize the production of the desired parent radionuclide. Such nucleosynthesis reactions are inadequate and can only produce low specific activity parents. By extracting only a small mass daughter component using the multi-column selective inversion generator, a large amount of isotopically enriched ions can be retained in the solution and easily recovered for later irradiation. Also important is the small volume of the solution that recovers the daughter radionuclide; made possible by the use of a small column and the logic of the multi-column selective reversal generator.
The method of the present invention is typically set up to operate substantially free of air or gas, thereby allowing better chromatographic performance. The presence of interstitial gas pockets results in the solution passing through the passage without flowing over, in or around the beads; rather, the solution passes through the passage without contact with the separation medium. In particular, air or gas passing through the separation medium can create a passage where less desired intimate contact between the solution and the separation medium can occur. Thus, each column used in the intended method is set up as a system for transporting and processing liquids.
Another advantage for such air-free or gas-systems is that there is no air or gas that must be sterilized by filtration through a sterile air filter. Thus, the components used in the intended method may have a less complex design than components that use a combination of air and liquid.
The usefulness of this generator technology is significant, and the versatility of the basic logic shown in FIG. 3 means that many types of radionuclides can be purified using the multi-column selective inversion generator concept. Table 1 below presents a list of radionuclides of interest for imaging or therapeutic nuclear medicine, along with exemplary solution conditions and chromatographic materials in their purification using a multi-column selective reversal generator. The inventory of radionuclides and separation conditions reported in Table 1 should not be construed as limiting, but rather how to use different parent / daughter pairs with completely different solution chemistries, ionic charges and charge densities in nuclear medicine applications. It is an example which can be separated and purified for use. As new separation media become available and interest in other radionuclides has increased, this multi-column selective reversal generator provides high chemical and radionuclide purity for use in diagnostic or therapeutic nuclear medicine. It can be easily adapted to provide a convenient route for reliable production of radionuclides.

^a 核医薬団体によって定義された医療上有用な放射性核種。[Bond et al., Ind. Eng. Chem. Res. (2000) 39:3130-3134]。
^b 数種の生産経路が多くの場合存在し、表中の経路は、核医薬において一般的に受入れられている経路である。
^c 汎用されている分離方法として、AIX = アニオン交換クロマトグラフィー；CIX = カチオン交換クロマトグラフィー；EXC = 抽出クロマトグラフィー；AOPE-EXC = 酸性有機リン抽出剤(acidic organophosphorus extractant)-EXC；NE-EXC = 中性有機抽出剤(neutral organic extractant)-EXC, MF-NE-EXC = 多官能性中性有機抽出剤(multifunctional neutral organic extractant)-EXC；ABEC = 水性2相抽出クロマトグラフィー(Aqueous Biphasic Extraction Chromatography)がある。
^d 保持担体としては、カルボキシレート類、ポリアミノカルボキシレート類、ある種の無機アニオン類、キレート化剤等がある。
^e 生理食塩水 = 生理学的食塩溶液。 Table 1

^a Medically useful radionuclide defined by ^a nuclear medicine organization. [Bond et al., Ind. Eng. Chem. Res. (2000) 39: 3130-3134].
^bSeveral production routes often exist, and the routes in the table are those generally accepted in nuclear medicine.
As separation ^{methods c} being generic, AIX = anion exchange chromatography; CIX = cation exchange chromatography; EXC = extraction chromatography; AOPE-EXC = acidic organophosphorus extractant (a cidic o rgano p hosphorus e xtractant) -EXC ; NE-EXC = neutral organic extractant (n eutral organic e xtractant) -EXC , MF-NE-EXC = polyfunctional neutral organic extractant (m ulti f unctional n eutral organic e xtractant) -EXC; ABEC = there aqueous two-phase extraction chromatography (A queous B iphasic E xtraction C hromatography).
^{Examples of d-} supporting carriers include carboxylates, polyaminocarboxylates, certain inorganic anions, and chelating agents.
^eSaline = physiological saline solution.

The intended method and system may use one or more separation media. The separation medium or media used in a given separation is occupied by the radionuclide to be separated, as is well known. Preferred separation media are typically in the form of beads or have a solid phase resin of consistent size and form, although sheet, web or fiber separation media may also be used.
One preferred solid phase separation media are commercially in the H ⁺ form, available Bio-Rad ^R 50W-X8 cation exchange resin from Bio-Rad Laboratories Inc., Hercules, CA. Other useful strongly acidic cation exchange medium, there is a Bio-Rad ^R AGMP-50 and Dowex ^R 50 W groups of the ion exchange resin and Amberlite ^R IR groups of the ion exchange resin available from St. Louis, Mo. Sigma Chemical Company. Anion exchange resin such as Bio-Rad ^R AGMP-1 and Dowex ^R 1 group of the anion exchange resin may also act as a separation medium.
Another resin that can be used in the process of the present invention is a styrene-divinylbenzene polymer matrix containing chemically bonded sulfonic acid, phosphonic acid and / or gem-diphosphonic acid functional groups. Such gem- diphosphate resins are commercially available under the trade name Diphonix ^R resin from Eichrom Technologies Inc., Cass Avenue 8205 S Illinois Darien. In the process of the present invention, Diphonix ^R resin is used in H ⁺ form. The characteristics and properties of Diphonix ^R resin are more fully described in US Pat. No. 5,539,003, US Pat. No. 5,449,462 and US Pat. No. 5,281,631.
TEVA ^™ resin containing a quaternary ammonium salt, especially trioctyl and tridecylmethylammonium chloride, adsorbed on a water-insoluble support that is inert to the components of the exchange composition is a tetravalent oxidation state. Is highly selective for ions having For example, the +4 valent thorium ions in a nitric acid solution, while binding to TEVA ^TM resin, actinium (Ac) and radium (Ra) ions (these valences, respectively, +3 +2 Are not substantially extracted by contact with this resin under the same conditions. TEVA ^™ resin is commercially available from Eichrom Technologies.

In the intended method, the second separation medium (ion exchange medium) contains a diphosphonic acid (DPA) ligand or group. Several types of DPA-containing substituted diphosphonic acids are known in the art and can be used in the present invention. One example diphosphonic acid ligand has the formula: CR ¹ R ² (PO ₃ R ₂ ) ₂
(Wherein, R is hydrogen (hydrido), C ₁ -C ₈ alkyl group, selected cations, and mixtures thereof; R ¹ is hydrogen or C ₁ -C ₂ alkyl group; R ² is a bond to hydrogen or polymer resin).
When R ² is a bond to a polymeric resin, the phosphorus-containing group is present at 1.0 to about 10 mmol / g dry mass of the copolymer, and its mmol / g value is based on the polymer where R ¹ is hydrogen. To do. Examples of exchange media containing diphosphonic acid ligands are described below.
One such exchange medium is an extraction chromatographic material containing a liquid diphosphonic acid extractant belonging to the group of diesterified methanediphosphonic acids, such as Dipex ^R resin, and di-2-ethylhexyl methanediphosphonic acid. is there. The extractant is adsorbed onto a substrate, such as inert Amberchrom ^R -CG71 (available from TosoHaas Inc., Pennsylvania Montgomery building) or hydrophobic silica relative to the mobile phase. In this extractant, R ¹ and R ² are H, one R is 2-ethylhexyl and the other is H.
Dipex ^R resin has high affinity for trivalent lanthanides, various trivalent, tetravalent and hexavalent actinides, and trivalent cations of preactinide ²²⁵ Ac, and radium cations and ²²⁵ It has been shown to have a lower affinity for certain decay products of Ac. These affinities have been shown even in the presence of complexing anions such as fluorides, oxalates and phosphates.

(式中、Rは、C₆〜C₁₈アルキル又はアリール、好ましくは2-エチル-1-ヘキサノールから誘導されたエステルである)。
好ましい化合物は、P,P'-ビス-2-(エチル)ヘキシルメタンジホスホン酸である。
上記活性成分ジホスホン酸エステルは、クロマトグラフィーカラムにおいて使用する当該技術において公知のようにして、メタノール、エタノール、アセトン、ジエチルエーテル、メチルエチルケトン、ヘキサン類又はトルエンのような低沸騰性有機溶媒と混合し、ガラスビーズ、ポリプロピレンビーズ、ポリエステルビーズ又はシリカゲルのような不活性支持体上にコーティーングし得る。ペンシルベニア州フィラデルフィアのRohm and Haas社から商業的に入手可能なAMBERLITE^Rのようなアクリル及びポリ芳香族樹脂も使用し得る。
Dipex^R樹脂の性質と特徴は、Horwitz等の米国特許第5,651,883号及びHorwitz等の米国特許第5,851,401号においてより十分に記載されている。Dipex^R樹脂は、Eichrom Technologies社から入手し得る。
もう１つの有用なイオン交換樹脂は、Diphosil^TM樹脂である。他のDPA樹脂と同様に、Diphosil^TM樹脂は、ビニリデンジホスホン酸によって調製したもののような複数のジェミナル置換ジホスホン酸リガンドを含有する。これらのリガンドは、シリカ粒子にグラフトさせた有機マトリックスに化学的に結合させる。Diphosil^TM樹脂は、Eichrom Technologies社から入手し得る。 A preferred Dipex ^R resin active ingredient is a liquid diphosphonic acid of the general formula:

Wherein R is an ester derived from C ₆ -C ₁₈ alkyl or aryl, preferably 2-ethyl-1-hexanol.
A preferred compound is P, P′-bis-2- (ethyl) hexylmethane diphosphonic acid.
The active ingredient diphosphonate is mixed with a low boiling organic solvent such as methanol, ethanol, acetone, diethyl ether, methyl ethyl ketone, hexanes or toluene as known in the art for use in chromatography columns, It can be coated on an inert support such as glass beads, polypropylene beads, polyester beads or silica gel. Acrylic and polyaromatic resins such as commercially available AMBERLITE ^R from Philadelphia, Pa. Rohm and Haas Company may also be used.
The properties and characteristics of Dipex ^R resin are more fully described in Horwitz et al. US Pat. No. 5,651,883 and Horwitz et al. US Pat. No. 5,851,401. Dipex ^R resin is available from Eichrom Technologies.
Another useful ion exchange resin is Diphosil ^™ resin. Like other DPA resins, Diphosil ^™ resins contain multiple geminal substituted diphosphonic acid ligands, such as those prepared with vinylidene diphosphonic acid. These ligands are chemically bonded to an organic matrix grafted onto silica particles. Diphosil ^™ resin is available from Eichrom Technologies.

最初に調製したときの意図するペンダントメチレンジホスホネートは、2個のC₁〜C₈ジアルキルホスホネートエステル基を典型的に含有する。これらエステルのC₁〜C₈アルキル基及び本明細書において述べた他のC₁〜C₈アルキル基の例としては、周知のように、メチル、エチル、プロピル、イソプロピル、ブチル、t-ブチル、ペンチル、シクロペンチル、ヘキシル、シクロヘキシル、4-メチルシクロペンチル、ヘプチル、オクチル、シクロオクチル、3-エチルシクロヘキシル等がある。イソプロピル基が好ましいR基である。R¹ C₁〜C₂アルキル基はメチル又はエチル基であり、R¹は最も好ましくは水素である。
調製後、上記アルキルエステル基を加水分解して、使用のために、上記式中のRが水素(プロトン)、Ca²⁺イオン、又はリチウム、ナトリウム若しくはカリウムイオンのようなアルカリ金属イオンであるようにする。
好ましくは、上記不溶性コポリマーは少なくとも2モル％の反応したビニルベンジルハライドを含有し、この割合は、より好ましくは、約10〜約95モル％である。上述の１種以上の反応したモノエチレン系不飽和モノマーは約2〜約85モル％で存在し、このモノマーは、好ましくは、少なくとも5モル％のスチレン、エチルスチレン、ビニルトルエン(メチルスチレン)及びビニルキシレンのような上記モノエチレン系不飽和芳香族モノマーを含む。 Yet another useful resin has pendant-CR ¹ (PO ₃ R ₂ ) ₂ groups attached by grafting to a pre-prepared water-insoluble copolymer; that is, the pendant phosphonate group is added after copolymer particle formation. Let In these polymers, R represents hydrogen (hydrido), an C ₁ -C ₈ alkyl group, cationic or mixtures thereof, R ¹ is hydrogen or C ₁ -C ₈ alkyl group. The intended pendant —CR ¹ (PO ₃ R ₂ ) ₂ group in this group of resins has the formula shown below. The particles also contain pendant aromatic sulfonate groups from 0 to about 5 mmol / g dry mass.

The intended pendant methylene diphosphonate when initially prepared typically contains two C ₁ -C ₈ dialkylphosphonate ester groups. Examples of C ₁ -C ₈ alkyl group and the other C ₁ -C ₈ alkyl groups mentioned herein, these esters, as is well known, methyl, ethyl, propyl, isopropyl, butyl, t- butyl, Examples include pentyl, cyclopentyl, hexyl, cyclohexyl, 4-methylcyclopentyl, heptyl, octyl, cyclooctyl, 3-ethylcyclohexyl and the like. An isopropyl group is a preferred R group. The R ¹ C ₁ -C ₂ alkyl group is a methyl or ethyl group, and R ¹ is most preferably hydrogen.
After preparation, the alkyl ester group is hydrolyzed so that for use, R in the above formula is hydrogen (proton), a Ca ²⁺ ion, or an alkali metal ion such as a lithium, sodium or potassium ion. To.
Preferably, the insoluble copolymer contains at least 2 mole percent reacted vinyl benzyl halide, and this proportion is more preferably from about 10 to about 95 mole percent. The one or more reacted monoethylenically unsaturated monomers described above are present in about 2 to about 85 mole percent, and this monomer is preferably at least 5 mole percent styrene, ethyl styrene, vinyl toluene (methyl styrene) and The monoethylenically unsaturated aromatic monomer such as vinyl xylene is included.

Useful insoluble copolymers also include reacted crosslinkers (crosslinkers). Also, the reacted crosslinkers useful in the present invention are quite varied. Useful examples of crosslinkers include divinylbenzene, trimethylolpropane triacrylate or trimethacrylate, erythritol tetraacrylate or tetramethacrylate, 3,4-dihydroxy-1,5-hexadiene and 2,4-dimethyl-1,5- Selected from the group consisting of hexadiene. Divinylbenzene is particularly preferred in the present invention. The amount of reacted crosslinker is that amount sufficient to achieve the desired insolubility. Typically, at least 0.3 mol% reacted crosslinker is present. The reacted cross-linking agent is preferably present at about 2 to about 20 mole percent.
These intended particles are multistage reaction products of a nucleophile such as CR ¹ (PO ₃ R ₂ ) ₂ ^— and a substrate that can be obtained by known methods. That is, CHR ¹ (PO ₃ R ₂ ) ₂ (R is preferably an alkyl group) is first formed into sodium or potassium metal, sodium hydride or organolithium compound, such as butyl lithium, or diphosphonate carbanion. React with any drug that can be made. The resulting carbanion is then reacted with a substrate which is an insoluble cross-linked copolymer as described above of one or more vinyl aliphatic, acrylic or aromatic compounds and polyvinyl aliphatic, acrylic or aromatic compounds such as divinylbenzene. . This copolymer comprises at least 2 mol%, preferably 10 to 95 mol% of a reacted halogenated derivative of a vinyl aromatic hydrocarbon such as a vinylbenzyl chloride group, about 2 to 85 mol% of a monovinyl aromatic such as styrene. It contains a hydrocarbon and at least 0.3 mol%, preferably 2-20 mol% of a polyvinyl aliphatic and / or aromatic crosslinker such as divinylbenzene.
The copolymer containing grafted methylene diphosphonate tetraalkyl ester groups in an amount corresponding to a dry weight of about 1.0 mmol / g, preferably 2-7 mmol / g, is preferably chlorosulfonic acid, concentrated sulfuric acid or Reacting with a sulfonating agent such as sulfur trioxide introduces a strongly acidic pendant aromatic sulfone group into the structure. The presence of sulfonate pendant groups gives the particles additional hydrophilic benefits, resulting in a surprising increase in the rate of cation complex formation without adversely affecting the observed selectivity.

Reaction of sulfonating agents with grafted copolymers containing methylene diphosphonate groups usually results in the recovery of the ester form of the recovered resin product by halohydrocarbons such as dichloromethane, ethylene dichloride, chloroform or 1,1,1-trichloroethane. When you do. This sulfonation reaction uses 0.5-20.0% by weight of chlorosulfonic acid in one of the halohydrocarbon solvents described above and ranges from about -25 ° C to about 50 ° C, preferably from about 10 ° C to about 30 ° C. Can be carried out at the following temperatures. The reaction is carried out by contacting the preswelled resin from 0 (non-swelled) to about 2 hours with the above sulfonated solution for 0.25 to 20 hours, preferably 0.5 to 2 hours.
After completion of the sulfonation reaction, the particles are separated from the liquid reaction medium by filtration, centrifugation, decantation and the like. This final second resin product is carefully washed with dioxane, water, 1 M NaOH, water, 1 M HCl and water and then air dried.
Treatment with the sulfonation reaction and water, also hydrolyze the phosphonate C ₁ -C ₈ alkyl ester group. Without sulfonation, the phosphonate ester can be hydrolyzed by reaction with an acid such as concentrated hydrochloric acid under reflux.
These intended particles contain both methylene diphosphonate and sulfonate groups as pendant functional groups bonded directly to carbon atoms of aromatic or acrylate or methacrylate units in the polymer matrix. The intended resin exhibits high affinity for a wide range of divalent, trivalent and multivalent cations over a wide range of pH values. At pH values below 1, the resin can switch from an ion exchange mechanism for cation removal to a bifunctional ion exchange / coordination mechanism based on the coordination ability of phosphoryl oxygen atoms. In that case, the sulfonic acid groups act to make the matrix more hydrophilic for rapid metal ion acceptance; thus, the methylene diphosphonate groups contribute to high selectivity. Further details on the production of this resin can be found in Trochimczuk et al., US Pat. No. 5,618,851.

Another particularly useful separation medium described in US Pat. No. 5,110,474 is referred to as Sr Resin and is available from Eichrom Technologies. In short, Sr resin contains an inert resin substrate on which a crown ether extractant solution dissolved in a liquid diluent is dispersed.
The diluent described above has (i) a high boiling point, i.e. a boiling point of about 170 ° C to 200 ° C, (ii) limited or no solubility in water, and (iii) about 0.5 to 6.0 M. Water is soluble and (iv) crown ether is a soluble substance. These diluents include alcohols, ketones, carboxylic acids and esters. Suitable alcohols include 1-octanol, which is most preferred, but 1-heptanol and 1-decanol may also be satisfactory. As carboxylic acid, in addition to heptanoic acid and hexanoic acid, octanoic acid is preferable. Examples of ketones include 2-hexanone and 4-methyl-2-pentanone, while esters include butyl acetate and pentyl acetate.
The macrocyclic polyether can be any dicyclohexano crown ether such as dicyclohexano-18-crown-6, dicyclohexano-21-crown-7 and dicyclohexano-24-crown-8. Preferred crown ethers have the formula: 4,4 ′ (5 ′) [(R, R ′) dicyclohexano] -18-crown-6, where R and R ′ are H and 1-12 carbons. One or more members selected from the group consisting of linear or branched alkyl containing Examples are methyl, propyl, isobutyl, t-butyl, hexyl and heptyl. Preferred ethers include dicyclohexano-18-crown-6 (DCH18C6) and bis-methylcyclohexano-18-crown-6 (DMeCH18C6). The most preferred ether is bis-4,4 ′ (5 ′)-[(t-butyl) cyclohexano] -18-crown-6 (Dt-BuCH18C6).
The amount of crown ether in the diluent can vary depending on the particular shape of the crown ether. For example, a concentration of about 0.1 to about 0.5 M of the most preferred t-butyl form (Dt-BuCH18C6) in the diluent is satisfactory, with about 0.2 M being most preferred. When using the hydrogen form, the concentration can vary from about 0.25 to about 0.5 M.

A preferred Sr resin has a crown ether coating layer on the surface, such as Dt-BuCH18C6 (20-25% by weight) dissolved in n-octanol (5-20% by weight), and 40% by weight using the extractant inactive resin substrate is a nonionic acrylic ester polymer beads resins such as Amberlite ^R XAD-7 having a load (60 to 70% by weight). [See Horwitz et al., Solvent Extr. Ion Exch., 10 (2): 31316 (1992)].
Further, (Again, available from Eichrom Technologies Inc.) Pb resin is related resins, it has been observed to be useful in purification of the ²¹² Pb in the production of ²¹² Bi accumulation. Pb resin has the same characteristics as Sr resin except that a higher molecular weight alcohol, i.e., isodecyl alcohol, is used in the production of Pb resin. [See Horwitz et al., Anal. Chim. Acta, 292: 26373 (1994)]. Pb resin has been observed to allow subsequent ²¹² Bi stripping from this resin, whereas ²¹² Pb has been observed to be strongly retained by Sr resin.
The improved Sr resin, also available from Eichrom Technologies, is somewhat more selective. The separation medium is referred to as Super Pb (Sr) ^TM selective resin, polymer resin (e.g., Amberchrom ^R CG 71), or from about 5 to about 50 dispersed on an inert porous support, such as silica particles Bis-4,4 ′ (5 ′)-[C ₃ -C ₈ -alkylcyclohexano] -18-crown-6 (about 10 ³ or more, usually about 10 ³ to about 10% by weight% of Dt-BuCH18C6 Includes free-flowing particles with a partition ratio between ⁶ n-octanol and 1 M nitric acid (showing D _crown = [crown _organic ] / [crown] _Ag )). This separation medium is free of diluent, in particular (i) is insoluble in water or has only limited (poor) solubility, and (ii) the substance present in the Sr resin. Does not contain a diluent that can dissolve the correct amount of water. See, for example, US Pat. No. 6,511,603 B1.
Also, the preferred cleaning and strip solutions used are selected based on the parent and daughter radionuclide and the desired use of the product. See Horwitz et al. US Pat. No. 5,854,968 and Dietz et al. US Pat. No. 5,863,439 for a specific description of this separation medium.

Yet another separation medium is particularly useful in the separation of chaotropic anions in aqueous solution. This separation medium is available from Eichrom Technologies under the trademark ABEC ^R and comprises particles having a plurality of covalently bonded —X— (CH ₂ CH ₂ O) _n —CH ₂ CH ₂ R groups. Wherein X is O, S, NH or N— (CH ₂ CH ₂ O) _m —R ^{3, where} m is a number having an average value of 0 to about 225, and n is about 15 to about A number having an average value of 225, R ³ is hydrogen, C ₁ -C ₂ alkyl, 2-hydroxyethyl or CH ₂ CH ₂ R); R is —OH, — (CH ₂ CH ₂ O ) _n - C ₁ -C ₁₀ hydrocarbyl ether having a molecular weight of up to about one-tenth part of the molecular weight, carboxylate, sulfonate, phosphonate and -NR ¹ R ² group (R ¹ and R ² are, individually, hydrogen, C ₂ -C ₃ hydroxyalkyl or C ₁ -C ₆ alkyl or -NR ¹ R ^{2 taken} together are 0 or 1 oxygen atom or 0 or 1 further nitrogen in the ring Forming a 5- or 6-membered cyclic amine having an atom). The separated particles have a percent CH ₂ O / mm ² particle surface area greater than about 8000 and less than about 1,000,000.
Examples of chaotropic anion, Br ^1-and I ^1-simple anions such as, and TcO ₄ ^1-, there are anionic groups such as ReO ₄ ^1-or IO ₃ ^1-. The chaotropic anion can also be a complex of metal cation and halide or a pseudohalide anion. A particularly useful separation that can be performed using this separation medium is the separation of ^99m TcO ₄ ¹⁻ from an aqueous solution that also contains the parent radionuclide ⁹⁹ MoO ₄ ²⁻ ions. ABEC the ^R separation medium and further details on the use thereof, U.S. Patent No. 5,603,834, can be found in Japanese Patent No. 5,707,525 and No. 5,888,397.
An example of a chelating resin is the material known as Chelex ^™ resin, available from Bio-Rad Laboratories, which contains multiple iminodiacetate ligands, and similar ligands can be added to 4% beaded agarose (Missouri). (Available from Sigma Chemical Co., St. Louis).
In a preferred method using separation medium beads, the column is packed with support beads containing the separation medium. As the solution is passed through the beads, the solution flows over, in and around the beads and is in intimate contact with the separation medium.

All acids were trace metal grade and all other chemicals were ACS reagent grade and were used when received. ²⁰⁷ Bi and ¹³³ Ba radioactive tracers were each evaporated to dryness twice in concentrated HNO ₃ and dissolved in 0.50 M HNO ₃ before use. Standard radiometric procedures were used globally and all count rates were corrected for background.
Extraction chromatography material was prepared using the general procedure previously disclosed. (See Horwitz et al., Anal. Chem., 63: 522-525 (1991)). In short, a 0.25 M tri-n-octylphosphine oxide (TOPO) solution (0.78 g) in n-dodecane was dissolved in about 25 mL of ethanol and 50-100 μm Amberchrom ^R − in about 25 mL of ethanol. Mixed with CG71 resin (3.03 g). The mixture was rotated on a rotary evaporator at room temperature for about 30 minutes, after which the ethanol was evaporated in vacuo. Resulting solids called TOPO resin, 20% of Amberchrom ^R -CG71 on in the n- dodecane 0.25 M TOPO (w / w) corresponding to the load. Modified TRPO Resin was prepared in a similar manner, but this material did not contain n-dodecane diluent and the dispersing solvent was methanol rather than ethanol. TRPO Resin is, 50 to 100 Cyanex ^R -923 (mixture of n- alkyl phosphine oxides) which is 40 percent loaded onto Amberchrom ^R -CG71 of μm and diphenyl (pentyl) - containing an equimolar mixture of phosphonates.
Percent solids in Bio-Rad ^R AGMP-50 cation exchange resin was determined by drying a part of wet resin was transferred and then 110 ° C. to constant weight is obtained in an oven in tared vial. Each gravimetric analysis was performed in triplicate to obtain a percent solids of 48.6 (± 0.3) percent. All resins were stored in tightly capped containers and were not exposed to air for any length of time to avoid percent solids changes.

(式中、A_o = 樹脂との接触前の溶液での計数率、A_f = 樹脂との接触後の溶液での計数率、V = 樹脂と接触した溶液の容量(mL)、m_R = 湿潤樹脂の質量(g)であり、％固形分は樹脂の乾燥質量への変換を示す)。 All dry weight distribution ratios (D _w ) were measured by radiometry by batch contact of the resin with the desired solution at 25 (± 2) ° C. The dry weight distribution ratio is defined as:

( _Where A _o = counting rate in solution before contact with resin, A _f = counting rate in solution after contact with resin, V = volume of solution in contact with resin (mL), m _R = The mass (g) of wet resin, where% solids indicates conversion to dry mass of resin).

Batch uptake test, add μL volume of ¹³³ Ba or ²⁰⁷ Bi in 0.50 M HNO ₃ to 1.2 mL of solution of interest, mix gently, and remove 100 μL aliquot for γ-count (A _o ) Carried out by. 1 mL of the remaining solution (V) was added to a known mass of wet resin (m _R ) and centrifuged for 1 minute. The mixture was then gently stirred for 30 minutes (to ensure that the resin was properly suspended in the solution), then centrifuged for 1 minute and stirred for an additional 30 minutes. After centrifuging for 1 minute to settle the resin, the solution was pipetted and filtered through a 0.45 μm PTFE filter to remove any remaining suspended resin particles. A 100 μL aliquot was then taken for γ counting (A _f ). All dry weight distribution ratios were accurate to a two-digit effective number.
A constant amount of TRPO Resin in 0.20 M HCl was slurried into a 1.2 mL Bio-Spin ^R disposable plastic chromatography column (Bio-Rad Laboratories) to give a 0.5 mL bed volume (BV). A porous plastic frit was placed on the floor to prevent its fragmentation during the eluate addition. The column was conditioned by eluting 3.0 mL (6 BV) of 0.20 M HCl, followed by gravity elution of 2.0 mL (4 BV) of 0.20 M HCl plus ¹³³ Ba and ²⁰⁷ Bi. Each column was then washed with 2.0 mL (4 BV) of 0.20 M HCl and ²⁰⁷ Bi was stripped using 2.0 mL (4 BV) of 1.0 M sodium acetate (NaOAc) in 0.20 M NaCl. Column eluates were collected in tared γ counting vials and all volumes were calculated by weighing using the respective solution densities.

1.0% NHO ₃ in 20-50 μm DipexR Resin {40% P, P'-bis (2-ethylhexyl) methane diphosphonic acid on Amberchrom-CG71, Eichrom Technologies, [Horwitz et al., React. Funct. Polymers, 33: 25-36 (1997)]} was slurried into a custom plastic chromatography column to give 0.16 mL of BV. A porous plastic frit was used to keep the resin in place during the chromatography operation, which was done using a custom automated low pressure chromatography apparatus. The column was conditioned by eluting 4.0 mL (25 BV) of 1.0 M HNO ₃ and then elution of 2.0 mL (12.5 BV) of 1.0 M HNO ₃ with ¹³³ Ba and ²⁰⁷ Bi added was approximately 0.25 mL / Performed at a flow rate of minutes. Each column was then washed with 2.0 mL (12.5 BV) of 1.0 M HNO ₃ and ²⁰⁷ Bi was stripped using 2.0 mL (12.5 BV) of 2.0 M HCl. Column eluates were collected in tared γ counting vials and all volumes were calculated by weighing using the respective solution densities.
As mentioned above, the use of high LET α- and β ¹ releasing radiation has great prospects in the treatment of micrometastatic carcinomas and solid tumor masses. [Whitlock et al., Ind. Eng. Chem. Res. 39: 3135-3139 (2000); Hassfjell et al., Chem. Rev. 101: 2019-2036 (2001); Imam, Int. J. Radiation Oncology Biol Phys. 51: 271-278 (2001); and McDevitt et al., Science 294: 1537-1540 (2001)]. One candidate α-emitter proposed for cancer therapy is ²¹² Bi generated as part of the uranium-232 ( ²³² U) decay chain shown in FIG. 4 [Whitlock et al., Ind. Eng Chem. Res. (2000) 39: 3135-3139 (2000); Hassfjell et al., Chem. Rev. 101: 2019-2036 (2001); and Imam, Int. J. Radiation Oncology Biol. Phys. 51: 271-278 (2001)].

Bismuth-212 currently has a relatively long-lived (i.e., 3.66 d) ²²⁴ Ra parent retained on the cation exchange resin, and ²¹² Bi is usually eluted by about 1-3 M HCl or a mixture of HCl and HI. The elution from the generator is obtained for use. [See Mirzadeh et al., J. Radioanal. Nucl. Chem. 203: 471-488 (1996) and Mirzadeh, Appl. Radiat. Isot. 49: 345-349 (1998)]. Radiolysis of the above cation exchange resin limits the useful deployment life of the ²¹² Bi generator to approximately 2 weeks [see Mirzadeh et al., J. Radioanal. Nucl. Chem. 203: 471-488 (1998)] A multi-column selective reversal generator may provide advantages in the purification of ²¹² Bi. The decay chain leading to ²¹² Bi also provides an interesting test basis for the multi-column selective reversal generator concept, and the examples detailed below target the development of a new ²¹² Bi generator.
A study of the radioactivity half-life shown in Figure 4 suggests that a solution of ²²⁴ Ra with t _1/2 = 3.66 days is well suited to act as a radionuclide source for use in nuclear pharmaceuticals. ing. ²¹² Bi can be extracted from this solution using a primary separation column that is selective for Bi (III) but can elute Ra (II), Po (IV) and Pb (II). . In this ²¹² Bi embodiment, the most dangerous radionuclide impurity is the relatively long-lived osteophilic ²²⁴ Ra parent, and ²¹² Pb (t _1/2 = 10.64 h) is somewhat less of a concern.

The behavior of Ra (II) can be deduced from tests using its lighter homologous element Ba (II), and this chemical analogy is used in the following description. FIG. 5 shows TOPO Resin, an extraction chromatography material containing 0.25 M tri-n-octylphosphine oxide (TOPO) in n-dodecane at 20% loading on 50-100 μm Amberchrom-CG71. shows a plot of D _w in Ba (II) and Bi (III) with respect to [HCl] of.
This plot suggests the potential of TOPO Resin in the separation of Bi (III) from Ra (II) as an extension from Ba (II) and its chemical similarity in 0.04-0.4 M HCl. D _w values less than 10 obtained from these batch contact tests essentially indicate no adsorption of a given analyte (i.e., Ba (II) and Ra (II) as an extension are chromatographic elutions). Note that it will not be substantially retained under the conditions). By operating in the chromatographic mode rather than in the batch mode used to obtain the data of FIG. 5, a DF greater than 10 ³ of Ba (II) (and Ra (II)) from Bi (III) was achieved. Can be done.
Further, FIG. 5, D _w in Bi (III) have also shown that decreases at both ends of the HCl concentration, which is greater HCl concentration or pH = 3 to 10 buffered strip solution than 1 M Suggest that it can act as an effective stripping agent. Due to the proposed in vivo use of radionuclides and the need for conjugation to biolocalization agents, values close to physiological pH are preferred, strongly acidic media inhibit conjugation reactions and It can be chemically eroded.

A chromatographic test was performed to evaluate the effectiveness of stripping at low acid concentrations by specifically stripping with sodium acetate (NaOAc) solution at pH = 6.5. The chromatographic separation of Ba (II) from Bi (III) using modified TRPO Resin (which is very close to the above phosphine oxide-containing TOPO Resin) is shown in FIG. 6, where NaOAc is near neutral. Check the principle of stripping Bi (III) from TRPO Resin using at pH.
FIG. 6 shows that Ba (II) elutes with the first free column volume of 0.20 M HCl loading solution (as expected at less than 10 D _w from FIG. 5) and about 2 bed volumes of 0.20 M HCl wash. After that, it stably decreases to the background value. A small amount of ²⁰⁷ Bi (III) is detected in the loading column eluate, but is not statistically significant at less than twice the background radiation level in the ²⁰⁷ Bi window. ¹³³ Ba (II) could not be detected in a strip solution containing 1.0 M NaOAc in 0.20 M NaCl, effectively removing more than 85% Bi (III) in approximately 2 bed volumes. This test shows that Bi (III) effectively separates from Ba (II) and reduces the acid concentration from modified TRPO and TOPO Resin from pH = 0.70 (at 0.20 M HCl) to pH = 6.5 (1.0 M NaOAc). Confirm that it can be stripped.
The chromatogram in FIG. 6 shows that the TRPO Resin gives a DF of Ba (II) (and Ra (II)) from Bi (III) of about 10 ³ and that this resin is effective in a multi-column selective reversal generator. It was shown that it could act as a primary separation column. However, in order to ensure a high purity product and to minimize the likelihood that the ²²⁴ Ra and ²¹² Pb parents will reach the patient, while maintaining ²²⁴ Ra (II) and ²¹² Pb (II), ²¹² Bi (III) A guard column that enables elution was developed.

FIG. 7 shows the dependence of Bi (III) incorporation on the macroporous sulfonic acid cation exchange resin for [Cl ¹⁻ ] at two different pH values. A Cl ^1- concentration of about 1 M purifies anionic chloro complexes of Bi (III) that are not retained by the cation exchange resin (eg, BiCl ₄ ¹⁻ , BiCl ₅ ^2−, etc.). As a result, the _Dw value of Bi (III) shown in FIG. 7 is quite low, indicating that there is little, if any, retention of the anionic chloro complex of Bi (III) under chromatographic conditions. . Ra (II) retention by sulfonic acid cation exchange resins in this pH range has been reported to be quite high [Massart, "Nuclear Science Series, Radiochemical Techniques: Cation-Exchange Techniques in Radiochemistry," NAS-NS 3113; National Academy of Sciences, (1971)], ²²⁴ Ra (II) does not elute from the cation exchange resin guard column, suggesting that it does not significantly contaminate the ²¹² Bi (III) eluate. .
It has been reported that the extraction of Pb (II) from solutions below 1 M HCl with neutral organophosphorus extractants similar to those used in TOPO and TRPO Resin of the primary separation column is quite low. [See Sekine et al., Solvent Extraction Chemistry, Marcel Dekker, New York (1977)]. Proposed cation exchange resin guard column of Figure 7 is based on the observation that Pb (II) is not formed enough to be any recognition of the [Cl ^1-] In anionic chloro complexes of less than 1 M, ²¹² Pb ( Provide further decontamination from II). Supporting this observation is that ²¹² Bi (III) substantially free of the immediate ²¹² Pb (II) parent can be eluted from the sulfonic acid cation exchange resin with 0.5 M HCl (i.e., Pb (II) It is a test result reporting that it is held in a cation exchange resin under conditions. [Hassfjell et al., Chem. Rev. 101: 2019-2036 (2001); Mirzadeh et al., J. Radioanal. Nucl. Chem. 203: 471-488 (1996); and Mirzadeh ,. Appl. Radiat. Isot 49: 345-349 (1998)]. The data shown in FIGS. 5 to 7 are multi-columns based on the primary separation column of neutral organophosphorus extractant from ²¹² Bi and its parent of ²²⁴ Ra and ²¹² Pb, together with literature data on Pb (II). It suggests that a selective reversal generator can be used to effectively separate.

FIG. 8 shows an alternative to the modified TRPO Resin primary separation column (FIG. 6) in the separation of ²¹² Bi (III) from ²²⁴ Ra (II) and ²¹² Pb (II). Dipex ^R resin is an extraction chromatographic material consisting of 40% loaded P, P′-bis (2-ethylhexyl) methanediphosphonic acid on 20-50 μm Amberchrom-CG71. [See Horwitz et al., React. Funct. Polymers 33: 25-36 (1997)]. FIG. 8 shows that Bi (III) is strongly retained from 1.0 MHNO ₃ by Dipex ^R resin, but Ba (II) is easily eluted. A statistically significant amount of ²⁰⁷ Bi (III) was not detected during loading and washing procedures, and ¹³³ Ba (II) levels reached background after 5 bed volumes with 1.0 M HNO ₃ washing. Stripping with 2.0 M HCl removes more than 93% of ²⁰⁷ Bi (III) with a minimum amount of ¹³³ Ba (II) in two bed volumes. The use of chelating ion exchange Dipex ^R resin in the primary separation column gives an overall DF greater than 10 ³ but minimizes the possibility of contamination of ²¹² Bi product with ²²⁴ Ra and ²¹² Pb This would still require the use of guard column chemicals as described above.

The patents, patent applications and articles cited herein are incorporated by reference. Use of the article “a” or “an” shall include one or more.
From the above description, numerous modifications and variations can be made without departing from the true spirit and scope of the novel concepts of the present invention. It should be understood or inferred that no limitation is intended in connection with the particular embodiment illustrated. All disclosures, such as modifications within the scope of the claims, are intended to be embraced by the claims.

Scheme modified from Bond et al., Ind. Eng. Chem. Res. (2000) 39: 3130-3134 showing the seven basic steps in the production of medically useful radionuclides and the purity and control conditions of these radionuclides. It is. Fig. 6 is a schematic diagram showing a conventional generator approach using upflow elution as developed in ^99m TC. Is a schematic representation of a comprehensive logic multi column selective inversion generator described in this specification, PSC denotes a primary separation column (P rimary S eparation C olumn) , GC guard column (G uard C olumn) Show. Multi column selective inversion generator main impurity in the development of for ²¹² Bi (desired radionuclide, radium and lead species may interfere with the medical use of ²¹² Bi) stressed, indicating the radioactive decay chain from ²³² U to ²⁰⁸ Pb . TOPO Resin primary separation column on the of Ba (II) [closed circles] and Bi (III) dry weight distribution ratio of the black squares, a graph plotting against [HCl] molar concentration D _w. Separation of Ba (II) [white squares] from Bi (III) [white circles] with TRPO resin using 0.20 M HCl as the pre-equilibration, loading and washing solution and 1.0 M NaOAc in 0.02 M NaCl as the strip solution Elution in minutes per milliliter of eluate passed through a 25 (± 2) ° C column during loading (0.75 to 4.75 BV), washing (4.75 to 8.75 BV) and stripping (8.75 to 12.25 BV) procedures It is a graph with respect to the bed volume (BV) of a liquid quantity. The horizontal dotted line indicates the background quantity. ¹³³ Ba (II) was not observed in the above 8.75 to 12.25 BV range after the end of flow correction. Bi (III) for molar concentration of [Cl ^1- ] in 1.0 M sodium acetate / sodium chloride solution at pH 6.5 [black square] versus 0.0122 M HCl solution at pH 1.9 [black circle] in a sulfonate cation exchange resin guard column It is a graph which shows _Dw value of. In the separation of Ba (II) [white squares] from Bi (III) [white circles] with Bipex ^R resin using 1.0 M HNO ₃ as the pre-equilibrated, loading and washing solution and 2.0 M HCl as the strip solution, 25 Eluent volume per minute of eluate passed through (± 2) ° C column during loading (1-12 BV), washing (12-24.5 BV) and stripping (24.5-37 BV) procedures. It is a graph with respect to floor capacity (BV). The horizontal dotted line indicates the background quantity. ²⁰⁷ Bi (III) was not detected during loading. The quantity from ¹³³ Ba (II) reached the background level after passing 30 BV.

Claims (20)

A method for producing a desired daughter radionuclide solution free of impurities, comprising the following steps:
(a) a parent-daughter aqueous solution containing a desired daughter radionuclide, a first separation medium having a high affinity for the desired daughter radionuclide and a low affinity for the parent and other daughter radionuclides; Contacting, wherein the first separation medium contains a phosphine oxide or diphosphonate group and the desired daughter and the parent radionuclide are present in the solution, differ (i) ionic charge, (ii) charge density Or (iii) having both, and maintaining the contact for a time sufficient for the desired daughter radionuclide to bind to the first separation medium, the desired daughter-carrying separation medium and the desired daughter-deficient parent-daughter solution. Generating
(b) removing the desired daughter deficient parent-daughter solution from the separation medium;
(c) stripping the desired daughter radionuclide from the desired daughter-carrying separation medium to produce a solution of the desired daughter radionuclide;
(e) contacting the desired daughter radionuclide solution with a second separation medium having a high affinity for the parent radionuclide and a low affinity for the desired daughter radionuclide, wherein the contact is carried out by the parent radionuclide. Maintaining a sufficient time to bind to the second separation medium to produce a desired daughter radionuclide solution free of impurities;
The manufacturing method characterized by including. The method of claim 1, wherein the desired daughter and the parent radionuclide have different ionic charges. The method of claim 1, wherein the desired daughter and the parent radionuclide have different charge densities. The method of claim 1, wherein the desired daughter and the parent radionuclide have both different ionic charges and different charge densities. The method of claim 1, wherein a decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity of the first separation medium under the contact condition is 10 ² or more. The method of claim 1, wherein a decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity of the second separation medium under the contact condition is 10 ² or more. A method for producing a desired daughter radionuclide solution free of impurities, comprising the following steps:
(a) preparing a parent-daughter radionuclide aqueous solution containing the desired daughter radionuclide;
(b) contacting the parent-daughter solution with a first separation medium having a high affinity for the desired daughter radionuclide and a low affinity for the parent and other daughter radionuclides, under contact conditions; The decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity of the first separation medium is 10 ² or more, the first separation medium contains a phosphine oxide or diphosphonate group, When the desired daughter and the parent radionuclide are present in the solution, they have a different (i) ionic charge, (ii) charge density, or (iii) both, and the contact is desired by the desired daughter radionuclide. Maintaining a sufficient time to bind to the first separation medium to produce a desired daughter-carrying separation medium and a desired daughter-deficient parent-daughter solution;
(c) removing the desired daughter deficient parent-daughter solution from the separation medium;
(d) stripping the desired daughter radionuclide from the desired daughter-carrying separation medium to produce a desired daughter radionuclide solution;
(e) contacting said second desired daughter radionuclide solution with a second separation medium having a high affinity for said parent radionuclide and a low affinity for said desired daughter radionuclide, said second under contact conditions; decontamination factor of the desired daughter radionuclide from the parent radionuclide impurities of the separation medium is to be a 10 ² or more and sufficient the contact to the parent radionuclide bound to said second separation medium Producing a desired daughter radionuclide solution free of impurities by maintaining for a period of time;
The manufacturing method characterized by including. The method of claim 7, wherein the total decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity in both the first and second separation media is between 10 ^{4 and} 10 ¹⁰ . 8. The method of claim 7, wherein the desired daughter and the parent radionuclide have different ionic charges. 8. The method of claim 7, wherein the desired daughter and the parent radionuclide have different charge densities. 8. The method of claim 7, wherein the desired daughter and the parent radionuclide have both different ionic charges and different charge densities. The desired daughter radionuclide is selected from the group consisting of ⁹⁰ Y, ^99m Tc, ¹⁰³ Pd, ¹¹¹ In, ¹²⁵ I, ¹⁸⁸ Re, ²⁰¹ Tl, ⁴⁷ Sc, ²¹² Bi, ²¹³ Bi, ²¹¹ At and ²²³ Ra. Item 8. The method according to Item 7. A method for producing a desired daughter radionuclide solution free of impurities, comprising the following steps:
(a) A desired daughter radionuclide selected from the group consisting of ⁹⁰ Y, ^99m Tc, ¹⁰³ Pd, ¹¹¹ In, ¹²⁵ I, ¹⁸⁸ Re, ²⁰¹ Tl, ⁴⁷ Sc, ²¹² Bi, ²¹³ Bi, ²¹¹ At and ²²³ Ra. Preparing a parent-daughter aqueous solution containing,
(b) contacting the parent-daughter solution with a first separation medium having a high affinity for the desired daughter radionuclide and a low affinity for the parent and other daughter radionuclides, under contact conditions; The decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity of the first separation medium is 10 ² or more, the first separation medium contains a phosphine oxide or diphosphonate group, When the desired daughter and the parent radionuclide are present in the solution, they have different ionic charges and the contact is maintained for a time sufficient for the desired daughter radionuclide to bind to the first separation medium. Producing a desired daughter-carrying separation medium and a desired daughter-deficient parent-daughter solution;
(c) removing the desired daughter deficient parent-daughter solution from the separation medium;
(d) stripping the desired daughter radionuclide from the desired daughter-carrying separation medium to produce a desired daughter radionuclide solution;
(e) contacting said second desired daughter radionuclide solution with a second separation medium having a high affinity for said parent radionuclide and a low affinity for said desired daughter radionuclide, said second under contact conditions; the desired daughter decontamination factor of radionuclides as is 10 ² or more, and sufficient time the contact with the parent radionuclide to bind to said second separation medium from the parent radionuclide impurities of the separation medium Maintaining a desired daughter radionuclide solution free of impurities;
The manufacturing method characterized by including. The method of claim 13, wherein the total decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity in both the first and second separation media is between 10 ^{4 and} 10 ¹⁰ . A method for producing a desired daughter radionuclide solution free of impurities, comprising the following steps:
(a) Contains a desired daughter radionuclide selected from the group consisting of ⁹⁰ Y, ^99m Tc, ¹⁰³ Pd, ¹¹¹ In, ¹²⁵ I, ¹⁸⁸ Re, ²⁰¹ Tl, ⁴⁷ Sc, ²¹² Bi, ²¹³ Bi, ²¹¹ At and ²²³ Ra Preparing a parent-daughter radionuclide aqueous solution,
(b) contacting the parent-daughter solution with a first separation medium having a high affinity for the desired daughter radionuclide and a low affinity for the parent and other daughter radionuclides, under contact conditions; The decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity of the first separation medium is 10 ² or more, the first separation medium contains a phosphine oxide or diphosphonate group, When the desired daughter and parent radionuclide are present in the solution, they have different charge densities and the contact is maintained for a time sufficient for the desired daughter radionuclide to bind to the first separation medium. Producing a desired daughter-carrying separation medium and a desired daughter-deficient parent-daughter solution;
(c) removing the desired daughter deficient parent-daughter solution from the separation medium;
(d) stripping the desired daughter radionuclide from the desired daughter-carrying separation medium to produce a desired daughter radionuclide solution;
(e) contacting the desired daughter radionuclide solution with a second separation medium having a high affinity for the parent radionuclide and a low affinity for the desired daughter radionuclide so that the second separation medium under contact conditions the decontamination factor of the desired daughter radionuclide from a parent radionuclide impurities as is 10 ² or more, and a sufficient time maintaining the contact with the parent radionuclide to bind to said second separation medium And producing a desired daughter radionuclide solution free of impurities. The method of claim 15, wherein the total decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity in both the first and second separation media is between 10 ^{4 and} 10 ¹⁰ . 16. The method of claim 15, wherein the desired daughter radionuclide is ²¹² Bi (III). 18. The method of claim 17, wherein one parent radionuclide is ²²⁴ Ra (II). A method for producing a desired daughter radionuclide solution free of impurities, comprising the following steps:
(a) A desired daughter radionuclide selected from the group consisting of ⁹⁰ Y, ^99m Tc, ¹⁰³ Pd, ¹¹¹ In, ¹²⁵ I, ¹⁸⁸ Re, ²⁰¹ Tl, ⁴⁷ Sc, ²¹² Bi, ²¹³ Bi, ²¹¹ At and ²²³ Ra. Preparing a parent-daughter radionuclide aqueous solution containing,
(b) contacting the parent-daughter solution with a first separation medium having a high affinity for the desired daughter radionuclide and a low affinity for the parent and other daughter radionuclides, under contact conditions; The decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity of the first separation medium is 10 ² or more, the first separation medium contains a phosphine oxide or diphosphonate group, When the desired daughter and the parent radionuclide are present in the solution, they have both different ionic charges and charge densities, and the contact is sufficient for the desired daughter radionuclide to bind to the first separation medium. Generating a desired daughter-carrying separation medium and a desired daughter-deficient parent-daughter solution,
(c) removing the desired daughter deficient parent-daughter solution from the separation medium;
(d) stripping the desired daughter radionuclide from the desired daughter-carrying separation medium to produce a desired daughter radionuclide solution;
(e) contacting said second desired daughter radionuclide solution with a second separation medium having a high affinity for said parent radionuclide and a low affinity for said desired daughter radionuclide, said second under contact conditions; The decontamination factor of the desired daughter radionuclide from the parent radionuclide impurity of the separation medium is greater than 10 ² , and the contact is sufficient for the parent radionuclide to bind to the second separation medium; Producing a desired daughter radionuclide solution free of impurities by maintaining for a period of time;
The manufacturing method characterized by including. The method of claim 17, wherein the total decontamination count of the desired daughter radionuclide from the parent radionuclide impurity in both the first and second separation media is between 10 ^{4 and} 10 ¹⁰ .

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