WO2010000714A2

WO2010000714A2 - Radioimmunoconjugate kit

Info

Publication number: WO2010000714A2
Application number: PCT/EP2009/058145
Authority: WO
Inventors: Roy H. Larsen; Jostein Dahle
Original assignee: Oslo Universitetssykehus Hf
Priority date: 2008-07-04
Filing date: 2009-06-30
Publication date: 2010-01-07
Also published as: WO2010000714A3

Abstract

The present invention relates to a kit for purification of a solution comprising larger sized radioactive molecules, which solution can be administrated directly into a patient, comprising a concentrator, a solution comprising such compounds and a washing solution, a method for purifying a solution comprising larger sized radioactive compounds, which solution can be administrated directly into a patient, and use of the kit according to the present invention to prepare a sterile solution comprising larger sized radioactive compounds having a radiochemical purity higher than 70%. Furthermore the invention concerns a sterile concentrator for purifying a solution comprising larger sized radioactive compounds.

Description

RADIOIMMUNOCONJUGATE KIT

FIELD OF THE INVENTION

The present invention relates to a concentrator for purifying a solution comprising larger sized radioactive compounds; a kit for purification of a solution comprising larger sized radioactive compounds, which solution can be administrated directly into a patient; and a method for purifying a solution comprising larger sized radioactive compounds, which solution can be administrated directly into a patient, as well as use of the kit according to the present invention to prepare a sterile solution comprising larger sized radioactive compounds having a radiochemical purity higher than 70%.

BACKGROUND OF THE INVENTION

Radiopharmaceuticals may be defined as pharmaceutical substances containing radioactive atoms within their structure and are administered to humans. They are used both for diagnostic and therapeutic purposes such as imaging, in vivo function, in vitro studies and therapeutic treatment. Depending on the isotope, labels may be chosen based on their suitability for targeting different cell types. For instance, gamma emitters are generally used for diagnostic purposes while alpha and beta emitters may be used for therapeutic purposes.

Radiolabeled larger sized radioactive molecules, such as radioimmunoconjugates, are usually made by coupling a bifunctional chelator to a protein or antibody, and then the protein construct is radiolabeled with a radioisotope via the bifunctional chelator. In some cases the ligand is radiolabled before conjugation to antibodies. Iodine-125 or ¹³¹I can be directly bound to proteins by e.g., Chloramine-T (p- toluene sulfonochloramide) (see Hunter, W. M., and Greenwood, F.C., Nature 194:495 (1962)) or Iodogen Reagent (l,3,4,6-tetrachloro-3alpha,6alpha- diphenylglucoluril) (see Salacinzki, P.R.P., et al., Anal.Biochem. 117:136 (1981)) mediated oxidation, which causes iodination of tyrosine residues in the protein. Despite the recognized utility of radiolabeled larger sized molecules and the encouraging clinical results, many patients are deprived of the benefit these radiolabeled molecules might provide because of the inherent difficulties in conducting both the radiolabeling and administration at a single location. Due to their radioactive properties and complex nature radiopharmaceuticals are often produced at a few central production facilities and then sent to hospitals for use. This can give rise to contamination of the product over time by dissociated radioactive daughter nuclides including different cations, anions and neutral molecules with different small molecular groups as well as naked cations and anions, depending on the radioactive product. The chemical concentration of the contaminations will be low but still to high for clinical use and thus necessitate extensive purification of the radioactive product to remove unbound radionuclides. Alternatively the radiopharmaceuticals might be prepared by the nuclear medicine unit at the hospital close to the end user to avoid the transport and storage for a prolonged time.

Usually radiolabeled larger sized radioactive molecules, such as radioimmunoconjugates, are purified by use of a gel filtration column (e.g. based on materials such as Sephadex G-25 (GE Healthcare) (Zalutsky MR. et al., Pharmacokinetics and tumor localization of ¹³¹I-labeled anti-tenascin monoclonal antibody 81c6 in patients with gliomas and other intracranial malignancies. Cancer Res. 49, 2807-13, 1989), 10 DG (Biorad, Laboratories, Hercules, CA, USA) (Michael R. McDevitt, Ronald D. Finn, Dangshe Ma, Steven M. Larson and David A. Scheinberg. Preparation of ^-Emitting ²¹³Bi-labeled Antibody Constructs for Clinical Use J Nucl Med 1999, 40, 1722-27). Other examples using gel filtration for purification of radiolabeled larger sized compounds are given, e.g., in WO 89/12631, p. 139, Example XXIV, where a labeled antibody was isolated by centrifugal gel filtration on Sephadex™ G-25 columns; in WO 93/21963, pp. 20- 21, Example 8, a ¹⁷⁷Lu labeled F(ab')₂ was purified by centrifugal gel filtration on a Sephadex G-25 column; in WO 85/01442, p. 23, lines 5 to 29, an ¹²⁵I radiolabeled antibody was purified on a Sephadex G-25 M prepacked PD-IO column and p. 23, line 30 to p. 24, line 4, describes how an ¹³¹I labeled protein was separated by gel filtration on a Sepahdex PD-IO column; in US 6,683,162, column 11, lines 39-45, an ²²⁵Ac radiolabeled protein was purified using a Econo-Pack 10 DG gel filtration column; M. R. McDevitt et al., Applied Radiation and Isotopes 57 (2002), p. 844, describes purification of an ²²⁵Ac-DOTA-IgG construct by size exclusion chromatography; J. Dahle et al., in Nuclear Medicine and Biology 33 (2006) p. 273 describes purification of the conjugate of 227^Th-p-isothiocyanato-benzyl-DOTA to rituximab by gel filtration on Econo-Pac 10 Dg columns; and M. Miederer et al., The Journal of Nuclear Medicine, VoI 45 (1), Jan. 2004, p. 130, describes purification of an ²²⁵Ac-HuM195 by size exclusion chromatography.

If there is a charge difference between the radiopharmaceutical and the contamination(s), ion exchange chromatography is a suitable separation method. Furthermore, methods for purifying radiolabeled antibodies combining ion exchange and size exclusion resins have been used (U.S. Pat. No. 4,454,106). HPLC has also been used for purifying radiolabeled antibodies. Although other methods have been suggested (Subramanian R., Method for purifying chelator conjugated compounds, US Patent No. 5,244,816; issued Sept. 14, 1993) size exclusion/gel filtration using a separation column has until now been the preferred method for purifying radiolabeled macromolecules.

As mentioned above gel filtration is used to remove daughter nuclides from radioimmunoconjugates before administration of a radiopharmaceutical to a patient. The procedure is performed in sterile lab facility with sterile hoods where the antibody solutions are added to open columns with the gel-exclusion stationary phase. The elution is made by gravity. Several fractions have to be collected and both the volume and the radioactivity have to be measured by a technician in order to calculate the concentration of the radioimmunoconjugate. This method is thus time consuming and laborious and requires dedicated sterile conditions and benches for clinical radiopharmaceuticals, such as radioimmunoconjugates, in addition to an experienced and specially trained staff. This is inconvenient for many nuclear medicine units and would add to the expenses for the end user.

A further limitation to the use of gel exclusion chromatography is the propensity of antibodies to bind non-specifically and irreversibly to the chromatographic separation material. The radioimmunoconjugates currently on the market are not made in a central facility but typically by the nuclear medicine department in the hospital. This way of providing radioimmunoconjugates has contributed to limiting the use of radioimmunotherapy to large hospitals with both oncology and nuclear medicine departments (New York Times, July 14, 2007).

The radiopharmaceuticals such as solutions containing larger sized radioactive molecules are complex compounds which due to their nature give rise to more severe and faster evolving impurities than similar non-radioactive compounds. The most concerning impurities are non-conjugated radionuclides which may origin from at least three different chemical reactions of the radioactive product (i) chemical instability of the conjugate or complex, (ii) radiolysis of the product, (iii) radioactive decay yielding daughter nuclides that detach from the conjugate. It is important to reduce the normal tissue burdens from any non-conjugated radionuclides present in the radiopharmaceutical solution so as not to expose the patient to free radioisotopes which might accumulate in the bone and other non- target organs.

The amount and concentration of the different impurities will increase relatively rapidly with time for radiopharmaceuticals depending on the isotopes and the pharmaceutical substances constituting the radiopharmaceutical as many radionuclides and radiometals have a short half-life, which if not days, often is a matter of hours. By way of example, when ²²⁷Th decays, free ²²³Ra and radioactive daughter nuclides are formed and when ²²⁵Ac decays, free ²²¹Fr and radioactive daughter nuclides are formed. Furthermore, products based on ²²⁷Th and ²²⁵Ac can not be stored for more than 1 day and 1 hour, respectively, before the products need to be purified before use. The above example illustrates the quick and still different behavior with respect to ingrowth of radioactive daughters from different isotopes. It also makes it evident that rapid purification enhances the specific activity necessary for a labeled immunoconjugate.

Even though there exist several well known methods for purification of radiopharmaceuticals that provide satisfying results and administrable products, these methods are also related with several problems. As mentioned above the previously used purification methods are very time consuming as the gel filtration procedure for antibody solutions is performed in open columns through a gel-exclusion stationary phase where the elution takes place by gravity. Several fractions are collected and checked for both volume and radioactivity to finally calculate the concentration. Such a procedure may take about one hour or more to perform for a skilled technician when pre-equilibration of the column and counting of the fractions are included. In addition time is required for sterile preparation of column, sterilizing the sterile hood and cleaning up and removal of unused radioactive eluate fractions etc.

Due to the complexity of the above described purification procedure, the radioactive nature of the product combined with working in sterile labs only especially trained staff can handle the existing purification methods.

The labs where the purification takes place must have sterile facilities so as to assure sterile conditions for the radioactive product to be administered. This may be inconvenient for many nuclear medicine units as it would require dedicated sterile benches and would add to the expenses for the end user.

Additional shortcomings for the traditional methods are the costly equipment which is used as well as difficulties with respect to contamination of the equipment with radioactive spills. There may further be problems relating to the amounts of radioactive, hazardous waste materials formed during the purification process, e.g., if dialyse is used.

Even though the use of radiolabeled antibodies has evolved from research to approved clinical applications during the last decade, methods of preparation and purification of the final product have not changed much. This can be exemplified by one of the commercial radioimmunoconjugates on the market, Zevalin (⁹⁰Y- tiuexetan-ibritumomab), which is injected into the patient without a purification step after radiolabeling. The compound is shipped as a kit to nuclear medicine units which radiolabel the antibody according to a standard procedure. After radiolabeling, its clinical usefulness is assayed by a simple thin layer chromatography method ([http://www. zevalin. com/pdf/Zevalin_PI_webs'ite.pdf] version as of April 9^th 2008). If not 95% purity is confirmed by a first or a repeated thin layer chromatography, the batch is actually discarded. This is disadvantageously both to the patient for whom the treatment is delayed and for the hospital which has to spend even more resources by preparing a new batch to treat the waiting patient.

The major problem with respect to purifying radiolabeled immunoconjugates is further illustrated by the fact that new methods for preparing such compounds have been developed where the purification step is totally omitted, see US Patent No. 6,994,840 in the name of Biogen IDEC Inc. This patent discloses methods and kits for radiolabeling proteins, peptides and ligands with radioisotopes, particularly yttrium-90, whereby sufficient purity, specific activity and binding affinity are achieved such that the radiolabeled compound may be directly administered to the patient without the need for further purification.

The fact that solutions containing larger sized radioactive compounds currently require purification before administration, unless they are prepared by a method which do not cause impurities shortly before use, has been and will continue to be a major obstacle in their availability to patients who could benefit from their use unless a simplified method for purification is presented that allows physicians to quickly, efficiently and safely administer such reagents with high purity. It is thus of crucial importance to be able to purify such a solution by a simple, fast, inexpensive, reliable and reproducible method at the site where the patient will receive his or her treatment. These benefits will make it possible to use radiopharmaceuticals produced both at the site as well as radiopharmaceutical produced off-site.

Furthermore, a simple purification step that would remove the larger fraction of non-conjugated radionuclide would be of most interest since this could reduce the probability of a batch being discarded and would in general improve the purity of the product, which is to be injected.

Hence, there is a need for a simple, fast, inexpensive and reproducible method for purification of a solution comprising larger sized radioactive compounds wherein the unconjugated radionuclides are removed from the radioactive product. Accordingly, the problem to be solved by the present invention is thus to purify a solution comprising larger sized radioactive compounds such that the purified material will have a quality, in terms of both radiochemical purity and sterility, which is acceptable for administration directly into a patient.

Advantageously, the purification according to the invention can be carried out in a standard hospital laboratory by nursing staff avoiding time consuming procedures and costly equipments.

SUMMARY OF THE INVENTION

Accordingly, the invention preferably seeks to overcome the above mentioned disadvantages and provides a simple, fast, inexpensive and reproducible method for purifying a solution comprising larger sized radioactive molecules, where the purified solution will have a quality, in terms of both radiochemical purity and sterility, acceptable for administration directly into a patient.

Therefore, it is an object of the present invention to provide a concentrator useful for separating unconjugated radionuclides and other small sized molecules from the larger sized radioactive compounds in order to provide a solution comprising larger sized radioactive compounds which solution is to be administered directly to a patient as a sterile solution. It is a further object of the invention to provide a kit comprising the concentrator for carrying out the purification as well as a method for carrying out the purification. Furthermore the invention concerns the use of the abovementioned kit to prepare a sterile solution comprising larger sized radioactive compounds having a radiochemical purity higher than at least 70%, preferably at least 95%.

Additionally, the concentrator and kit including said concentrator are easy to dispose after use and produce minimal amounts of radioactive hazardous waste.

Thus, the above described objects and several other objects are intended to be obtained in a first aspect of the invention by providing a concentrator for removing or replacing at least unconjugated radionuclides in a solution comprising larger sized radioactive compounds without applying centrifugation, pressure or vacuum, the concentrator comprising a concentration chamber for receiving the solution, an absorption chamber for receiving at least the unconjugated radionuclides provided with absorbent material, the concentration and the absorption chambers being connected through a permeable membrane, the concentrator being a sterile and pyrogen-free closed unit having an opening for transferring and withdrawing solution sealed with a syringe-penetrable cap; and being equipped with a pressure compensation mechanism.

Different devices for concentrating and/or purifying non-radioactive macromolecules in solutions are known from, e.g., US 6,372,144, US 6,375,855 and US 6,837,995. Both US 6,375,855 and US 6,837,995 make use of an external driving force in one way or another to separate the macromolecules from the impurities while US 6,372,144 does not apply an external force. Other examples are, e.g., US 2002/0119095 Al, p. 27 section [0222], column 2, lines 2-3 using a Vivapore concentration device to concentrate unlabeled p97 protein; WO 89/12631, p. 139, Example XXIV, using a Centricon™ concentrator for pre- concentrating the unlabeled antibody before labeling; WO 2006/074397, Example 5, p. 116 describes concentration of an unlabeled ddCC49 molecule by use of an Amicon concentrator fitted with a YM30 membrane (Millipore); finally, the "Product Information Sheet and Operating Instructions Vivapore 2, 5, 10/20" [online] September 2007 (2007-09), VIVAPRODUCTS, Littleton, US, XP002507495 retrieved from internet: http:/www.vivascience-us.com/data_she ets/Vivapore_products.pdf> describes the use of the Vivapore device for purification of several types of larger sized molecules but is, totally, silent about the use of the device for purification of larger sized radioactive molecules, removal of daughter nuclides, etc.

The system described is US 6,372,144 is the purification device which has most in common with the present concentrator unit and is hereby incorporated by reference. The prior art devices are, however, not applicable for the preparation of radiopharmaceuticals, i.e., solutions comprising larger sized radioactive compounds, to be injected directly into a patient due to the fact that these devices are open to the ambient environment and that no precautions or steps have been taken to make and keep them sterile. As is well known by the person skilled in the art pharmaceutical procedures require among others that injectables are sterile and non-pyrogenic. This problem has been solved by the present invention in that the concentrator is sterile and pyrogen-free and has been sealed off from the environment with a rubber-cap being penetrable to syringe needles.

Another object of the invention is to provide a kit for purification of a solution comprising larger sized radioactive compounds, which solution can be administrated directly into a patient, comprising

• a concentrator for removing or replacing unconjugated radionuclides in a solution comprising larger sized radioactive compounds without applying centrifugation, pressure or vacuum, comprising a concentration chamber for receiving the solution, an absorption chamber for receiving at least the unconjugated radionuclides provided with absorbent material, the concentration and the absorption chambers being connected through a permeable membrane, where the concentrator is a sterile and pyrogen-free closed unit having an opening for transferring and withdrawing solution sealed with a syringe-penetrable cap;

• a solution containing larger sized radioactive compounds; and

• a washing solution.

The concentrator of the kit may additionally be equipped with a pressure compensation mechanism.

A further object of the invention is to provide a method for purifying a solution comprising larger sized radioactive compounds, which solution can be administrated directly into a patient, the method comprising the following steps - injecting a solution comprising larger sized radioactive compounds into the concentration chamber of the sterile, pre-washed concentrator of the invention;

- waiting for a predetermined time < 3 hours, more preferably 1 hour;

- withdrawing the concentrated solution comprising larger sized radioactive compounds from the concentration chamber into a sterile syringe containing washing solution;

- injecting a volume of the diluted washing solution comprising larger sized radioactive compounds into the concentration chamber of concentrator again; and

- withdraw the whole volume from the concentration chamber into the same syringe again. Still another object of the invention concerns use of a kit as disclosed above to prepare a sterile solution comprising larger sized radioactive compounds having a radiochemical purity of at least 70%, such as e.g. 71-75%, 71-80%, of at least 80%, such as e.g. 82%, 81-90%, 84-87%, 85%, of at least 90%, such as e.g. 91-94%, 91-93%, and preferably higher than 95% as measured by a gamma spectrometer.

The individual aspects of the present invention may each be combined with any of the other aspects. These and other aspects of the invention will be apparent from the following description with reference to the described embodiments.

BRIEF DESCRIPTION OF THE FIGURES

The figures show one way of implementing the present invention and is not to be construed as being limiting to other possible embodiments falling within the scope of the attached claim set.

Figures IA and IB showing side and front views, respectively, of the concentrator according to the present invention. Figure 1C showing side and front views, respectively, of the concentrator according to the present invention used in Example 8.

Figure 2 illustrates in a step-by-step manner how the kit of the present invention is used to purify a solution comprising larger sized radioactive compounds.

Figure 3 shows the average recovery Of ²²⁷Th-IgG and ²²³Ra in Example 1 after purification by use of a Vivapore 5 concentrator.

The present invention will now be described in more detail in the following.

DETAILED DESCRIPTION OF THE INVENTION

As is well known in the art of radiopharmacy, great interest and much effort have been put into the development of new radiopharmaceuticals, especially the generation of tumor-specific agents having a binding specificity for a tumor- specific antigen, such as radioimmunoconjugates. However, methods for purifying such compounds have not experienced the same development. Hence, the medical society has been depending on the same purification methods for years. The presently used purification methods are as discussed above very time consuming, require special facilities and well trained staff, all of which in practice will be a hurdle for the use of the new radiopharmaceuticals. The drawbacks of the former used methods can be summarized as shown in the following Table 1 : Table 1. Methods for purifying radiolabeled macromolecules

In some ways larger sized radioactive compounds behave as other larger sized compounds but additionally they create major difficulties not seen with nonradioactive compounds. Firstly, in addition to chemical instability of the conjugate or complex, radionuclides may origin from radioactive decay yielding daughter nuclides that detach from the conjugate. Secondly, radiolysis of the product might give radiolytic damage and breakdown of the compound. The mechanism for radiation damage has been attributed to the generation of free radicals (Pizzarello. 1975. Direct and indirect action. In : Pizzarello and Witcofski, eds. Basic Radiation Biology, 2.sup.nd ed. Philadelphia : Lea & Febger, pp. 20 29). In fact some radioisotopes, especially alpha and beta emitters, generate energy at a level which can break most chemical bounds. The purification of larger sized radioactive compounds differ from purification of other larger sized compounds also in that they normally have very low concentrations (nM) of the radionuclides which have to be removed. The chemical concentrations needed to produce biological effects are often very low for both radiopharmaceuticals and radioactive impurities, (e.g ., radioiodine released from iodinated antibodies may damage the thyroid if no protective measures are taken). Faced with these problems it is not obvious that methods that work well for purification of non-radioactive compounds can be used for radioactive compounds. The purification method has to be relatively fast to avoid formation of new contaminations in the purified product. Additionally, too high concentrations of the radioactive product should be omitted since this will increase the risk of radiolysis.

Other problems that have been seen with some of the purification techniques according to the prior art are that there is too much loss of the final product during purification, that the results are not reproducible and generate too much waste. The amount of waste, for example washing solutions, should be kept to a minimum both due to environmental consideration and also expenses involved in handling radioactive wastes. A simplified procedure could also reduce personnel exposure as well as the personnel's risk of being accidentally contaminated during performance of the procedure.

By way of example dialysis of a solution containing a ²²⁷Th~iabeied radioimmunoconjugate and unconjugated daughter nuclide ^Ra was performed, see the reference example m the experimental part. Dialysis was earned out agamst PBS or sterile water with volumes of least 10 times that of the sample, Although dialyses were tested over several hours, a sufficient purification of free ^/~'¹Ra from that of ^''Th-laheled antibody was not achieved,

A further problem to be solved by the present invention arises because it is desirable to administer the purified radioactive product directly to a patient without any additional loss of time.

The present invention overcomes the above mentioned problems and provides a simple, fast, inexpressive and reproducible method for purifying solutions comprising larger sized radioactive compounds. These purified solutions will be sterile and can be administered by injection directly into a patient. Thus, an aspect of the invention provides a concentrator for removing or replacing at least unconjugated radionuclides in a solution comprising larger sized radioactive compounds without applying centrifugation, pressure or vacuum, comprising a concentration chamber for receiving the solution, an absorption chamber for receiving at least the unconjugated radionuclides, where the absorption chamber is provided with absorbent material, and the concentration and the absorption chambers are connected through a permeable membrane, and wherein the concentrator is a sterile and pyrogen-free closed unit having an opening for transferring and withdrawing solution is sealed with a syringe-penetrable cap; and being equipped with a pressure compensation mechanism.

Larger sized compounds are according to the present invention considered to be molecules having a size above 10,000 Daltons, such as antibodies, including monoclonal and polyclonal antibodies, fragments and constructs of antibodies, nanocompounds including liposomes with or without surface ligands, and nanoparticles made from various new materials. Antibodies (also known as immunoglobulins) are gamma globulin proteins found in blood and other body fluids of vertebrates having a molecular weight of about 150 kDa, and are used by the immune system to identify and neutralize foreign objects, such as bacteria and viruses. Examples of antibodies are rituximab, trastuzumab, pertuzumab, ertumaxomab. Monoclonal antibodies are usually prepared by the hybridoma technology invented by Georges Kohler, Cesar Milstein, and Niels Kaj Jerne approximately 35 years ago. Liposomes are spherical vesicles formed by bilayer membranes composed of naturally-derived phospholipids with mixed lipid chains like egg phosphatidylethanolamine, or of pure surfactant components like DOPE (dioleoylphosphatidylethanolamine) having a size above 20 nm. The term nanoparticles is generally used to refer to small particles with all three dimensions less than 100 nm.

Alternatively, larger sized compounds according to the present invention may be defined as compounds or molecules that due to their size are unable to be transported across the permeable membrane connecting the concentration chamber and the absorption chamber. The term "larger sized radioactive compounds" refers to radioactively labeled compounds having a size above 10,000 Daltons, i.e. larger sized substances containing radioactive atoms or radionuclides within their structure. A radionuclide, also referred to as radioactive isotope or radioisotope, is an atom with an unstable nucleus characterized by having excess energy which is available to be imparted either to a newly-created radiation particle within the nucleus, or else to an atomic electron. In this process the radionuclide undergoes radioactive decay and emits gamma rays and/or subatomic particles.

The radioactive atoms or radionuclides bound to the larger sized compounds as well as radionuclides which are to be removed during purification by the present invention includes gamma-emitters, X-ray emitters, auger emitters, beta-emitters and alpha-emitters etc, i.e., all types of radionuclides useful for either imaging, therapeutic uses or measurement of physiological parameters etc. To exemplify the range of elements and isotopes that can be used, suitable candidates include, but are not limited to, ³²P, ⁴⁷Sc, ⁶⁷Cu, ⁷⁷As, ¹⁰⁵Rh, ¹⁰⁹Pd, ¹¹¹Ag, ¹⁴²Pr, ¹⁴⁹Pm, ¹⁵⁹Gd, ¹⁶⁶Ho, ¹⁷⁷Lu, ¹⁸⁶Re, ¹⁸⁸Re, ¹⁹⁴Ir, ¹⁹⁹Au, ⁸⁹Sr, ⁹⁰Y, ¹³¹I, ¹⁵³Sm, ²¹¹At, ²¹²Pb, ²¹²Bi, ²¹³Bi, ²²³Ra, ²²⁴Ra, ²²⁵Ra, ²²⁵Ac, ²²⁷Th, ²²⁸Th, ¹²³I, ¹²⁴I, ¹²⁵I, ¹⁵O, ¹³N, ¹¹C, ¹⁸F, ^99mTc, ¹¹¹In.

Typical examples of larger sized radioactive compounds that can be purified by the present invention are ⁹⁰Y-tiuexetan-ibritumomab (Zevalin™), ¹³¹I- tositumomab (Bexxar™), ²²⁷Th-labeled and ²²⁵Ac-labeled targeting molecules including antibodies and antibody fragments.

The term "unconjugated radionuclides" or "non-conjugated radionuclides" refers to radionuclides as mentioned above which are not bound or attached to the larger sized radioactive compounds. Radionuclides which are to be removed by the present invention comprise different chemical forms including cations, anions and neutral species formed with fragments or small molecular groups having a low molecular weight (i.e., < 2000 Daltons) of the larger sized compounds, as well as naked cations or anions of the radionuclide.

The term "patient", as used in the present invention, means a subject such as a human being or animal, preferably a human, which is to be administered, by any suitable rout of administration, a solution containing larger sized radioactive compounds. The rout of administration might be parenteral by injection or infusion such as intravenous, intramuscular, intrathecal; or, intravesical infusion, the preferred rout of administration is by injection.

As used in the present application the term "sterile" means that a device, surface or solution or microenvironment are made free of infectious microorganisms. The term "pyrogen-free" means that a device, surface, solution, etc. is free from pyrogens, i.e. endotoxins or exotoxins.

The term "administered directly", as used in the present invention, means that the solution prepared by use of the concentrator, kit or method according to the invention can be taken from the concentrator and administered without any further purification, e.g. sterile filtration, etc.

The concentrator for removing or replacing unconjugated radionuclides in a solution comprising larger sized radioactive compounds can be used without applying any external force such as centrifugation, pressure or vacuum which are often used in different standard methods for purification.

The sterile concentrator comprises a concentration chamber for receiving the solution comprising larger sized radioactive compounds and the impurities, i.e., unconjugated radionuclides, an absorption chamber for receiving at least the unconjugated radionuclides, where the absorption chamber is provided with absorbent material, the concentration and the absorption chambers being connected through a permeable membrane.

Figures IA, IB and 1C show side and front views of an embodiment of the concentrator. The concentrator 1 has a concentration chamber 2 for receiving the solution and an absorption chamber 3 for receiving at least the unconjugated radionuclides. The concentration and the absorption chambers are connected through a permeable membrane 4.

The permeable membrane 4 is made of a material allowing liquid and small molecules to pass through without need for vacuum, pressure or centrifugation, i.e., a size exclusion membrane. Suitable membranes are normal ultrafiltration or microporous membranes being highly hydrophilic and include cellulose triacetate, regenerated cellulose, polyetheresulfone, nylon, etc., preferably polyethersulfone. The cut-off of the permeable membrane should preferably be 7,500 MWCO or 30,000 MWCO. The membrane 4 should be formulated to provide low protein binding and high filtration speed.

The larger sized radioactive compounds will be concentrated inside the concentration chamber while a large fraction of any free radionuclide and small molecules will follow the vehicle solution through the membrane into the absorption chamber.

The absorption chamber 3 may be equipped with a liquid-absorbent material (not shown) which capture solvent and small sized compounds including non- conjugated radionuclides) crossing the permeable membrane 4. The absorbent material can be an absorbent cellulose pad mounted beneath the permeable membrane to draw solvents and radionuclides through the membrane. Other examples of suitable absorbent materials are silica gel and zeolites, hygroscopic salts etc. which preferably do not need heat activation to be an absorbent. Alternatively, the absorption chamber 3 may contain an ion exchange resin which can bind unconjugated ionic radionuclides. Depending on the radionuclides to be captured by the ion exchanger the ion exchanger can be cationic or anionic or a mixture of both.

The concentrator can be made sterile and pyrogen-free or non-pyrogenic by well known methods in the art (Disinfection, Sterilization and Preservation. Editor: Seymour S. Block. Lippincott, Williams & Wilkins, 2000). By way of example the concentrator can be sterilized by use of methods like autoclaving, gamma or X-ray irradiation etc. To keep the device sterile and pyrogen-free, the opening 5 of the concentrator is capped or sealed off with a cap 6 penetrable to syringe needles allowing the radiopharmaceutical liquid to be transferred, and later in a purified form, withdrawn from the concentration chamber. It should be mentioned that the absorption chamber does not necessary need to be kept sterile during use as the solution which is to be administered to the patient is kept in the concentration chamber of the concentrator. Thus, only the concentration chamber needs to be kept sterile during the purification process. The cap 6, as shown in Figures IA or IB, can be a rubber cap or a rubber cone 61, as shown in Figure 1C, and applicable materials are the same as those used for sterile medicine bottles, e.g., rubber, Teflon etc.. Furthermore the cap 6 or cone 61 can be protected by an outer cap (not shown) such as for example a metal cover which can be partly or totally removed before use.

The volume capacity of the concentrator 1 should fit the volumes of injectables, i.e. the solutions of larger sized compounds, to be purified. The volume capacity of the concentrator should be in the range from 0.1 - 999 ml, such as from , e.g. 0,1-5 ml, 5-10 ml, 0,1-100 ml, 50-100 ml, 100-250 ml, 200-900 ml, etc,, preferably from 1 - 50 ml and most preferred the volume capacity of the concentrator should allow several ml of radiopharmaceutical liquid to be loaded into the concentration chamber 2.

During use, the solution is transferred to and withdrawn from the concentration chamber 2, typically by use of a pipette or a syringe 20 inserted through an opening 5 in the concentrator providing access to the concentration chamber. The more detailed use of the concentrator will be described later in reference to Figure 2.

When volumes exceeding about 20% of the total volume of the concentrator are injected or withdrawn from the concentrator, it may be necessary to even out the increased or decreased air-pressure that may arise. Accordingly, the concentrator may include a pressure compensation mechanism. To keep the concentrator sterile, the pressure compensation mechanism can be equipped with a sterile filter.

In one embodiment, such mechanism can be made by, upon use, inserting a syringe needle with a sterile filter through the cap 6 of opening 5 located at the top of the concentrator 1, this is illustrated by the pressure compensation mechanism 21 in Figure IB.

In another embodiment, shown in Figure IA, the pressure compensation mechanism is a capillary 7 providing fluid access to the concentrator. Here, a capillary is a tube or opening of small internal diameter that holds liquid at capillary action. Such capillary could provide permanent fluidic access to the concentrator 1 and liquids, gels or powder materials would not escape through such capillary under normal conditions. Such capillary can be a small opening in, or a tube integrated in, the concentrator. A syringe needle, as shown in Figure 1C, may potentially (depending on dimensions and material selection) provide such capillary. In the embodiment, shown in Figure 1C, the pressure compensation mechanism is provided by plugging a hole, made on top of the absorption chamber of the concentrator, with a rubber cone 71 and then a sterile needle 7 is pinned through the cone. Furthermore, the capillary 7 may be equipped with a sterile filter.

In another embodiment (not shown), the pressure compensation mechanism is a two-way valve that allows flow of fluid only when the pressure inside the concentrator is higher or lower that the surrounding pressure. When not in use, e.g. during transport, the pressure in the concentrator will be equal to or close to that of the surroundings, and such valve will keep the content of the concentrator sealed. The capillary 7 may potentially (depending on dimensions and material selection) provide such two-way valve. Other implementations may be mechanical valves or valves formed by shaping of plastic materials such as those known from plastic bottles for e.g. honey or ketchup, but miniaturized.

To obtain reproducible results and high yields of the radiopharmaceutical the concentrator should be pre-washed to avoid protein binding and product absorption. Useful pre-wash solutions will typically contain proteins (e.g., albumins, immunoglobulins, etc.) and suitable pre-wash solutions are for example saline with 5% human serum albumin (HSA) or phosphate buffered saline (PBS) with 4% fetal calf serum (FCS). For human use human proteins should preferably be used in the pre-washing solution. For veterinary use proteins from the same species as the one receiving treatment should preferably be used. Such solutions must in general be compatible with in vivo injections. If desired, such solutions could also be used to wash the concentrator after purification of the solutions containing larger sized radioactive compounds. The concentrator may also be delivered as a ready-to-use concentrator where the concentrator has been pre-washed or equilibrated with the pre-wash solution by the manufacturer. Such ready-to-use concentrator can in one embodiment still contain the pre-wash solution, added by the manufacturer, so that is has to be emptied by the end-user. Said ready-to-use concentrators will have a shelf life of about 1-2 weeks. In an alternative embodiment the concentrator is pre-washed and then emptied by the manufacturer.

The concentrator of the present invention may in one embodiment be equipped with a pocket for the concentrate in the bottom of the concentration chamber to avoid concentration to dryness, thus providing a constant final volume of the solution to be purified. This pocket can be separated from the absorption chamber by means of an impermeable separation wall.

The container constituting the concentrator can be made of one or more waterproof materials with limited protein binding properties, such as glass, plastic or metal.

In an alternative embodiment, especially adapted for highly concentrated larger sized radioactive compounds, the concentrator can be provided with a constant volume diafiltration system e.g. as described on www.sartorius- stedim.com/fϋeadmin/sartoriυsjdf/bioLechJab/DataVPoreSLUΘOΘS-e.pdr. In this case, new buffer solution from a reservoir replaces the original solvent at a rate equal to the speed of filtration. This would be particularly useful for highly radioactive products, where a strong concentration could cause radiolysis of the product because radiation dose rate increases with decreasing volume of the sample. Another advantage is that the protein concentration can be kept low with such procedures because the flow out of concentration chamber will be balanced by the flow of buffer into the concentration chamber. This may be important if the starting solution is relatively concentrated because filtration rates may be slowed down if the protein concentration exceeds 2% (w/w).

In another embodiment the concentrator is placed inside an outer vial of a suitable material, e.g., glass or plastic. The outer vial is sealed with an air-tight rubber cap. Preferentially the outer vial fits exactly both in length and diameter so the concentrator is kept in place by the rubber cap. A pressure regulator, e.g., a syringe tip with a sterile filter is inserted into the rubber cap. The solution in need of purification is transferred to the concentrator by inserting a syringe tip through the rubber cap and injecting the solution into the concentration chamber of the concentrator. The same would apply to washing solution(s). Withdrawal of solutions is performed after inserting a syringe tip through the rubber cap. Thus the concentrator can be sealed from air and kept sterile and the solutions can be kept sterile during the purification and washing procedures. The concentrator will be sterilized and treated against pyrogens before insertion into the pre-sterilized and pre-pyrogen-treated outer vial and rubber cap. Alternatively, a pre-assembled unit may be sterilized by, e.g., irradiation and/or temperature treatment.

The concentrator of the present invention should be disposed as a radioactive material after use.

The solution comprising the larger sized compounds to be purified is injected into the sterile concentrator trough the rubber cap or cone using a sterile syringe. The larger sized radioactive compounds will be concentrated in the concentration chamber while a large fraction of radionuclides will follow the vehicle through the membrane and be absorbed in a liquid absorbent material in the absorption chamber. After a period of less than 3 hours, preferably 15-120 minutes, typically 60 minutes, the initial volume has been concentrated to 1-20% (v/v), while simultaneously removing 99-80% (v/v) of the liquid and small molecular compounds. As an example of such a concentrator the commercially available Vivapore 5 concentrator (Sartorius Stedim Biotech, Aubagne, France) has been used to remove impurities in the form of non-conjugated radionuclides. However, the concentrator of the present invention differs from the Vivapore 5 concentrator in that it is a closed unit that is sterile and pyrogen-free as described above.

Another aspect of the invention relates to a kit for purification of a solution comprising larger sized radioactive compounds, which solution can be administrated directly into a patient, comprising

- a concentrator for removing or replacing unconjugated radionuclides in a solution comprising larger sized radioactive compounds without applying centrifugation, pressure or vacuum, comprising a concentration chamber for receiving the solution, an absorption chamber for receiving at least the unconjugated radionuclides provided with absorbent material, the concentration and the absorption chambers being connected through a permeable membrane, where the concentrator is a sterile and pyrogen-free unit having an opening for transferring and withdrawing solution sealed with a syringe-penetrable cap;

- a solution containing larger sized radioactive compounds; and

- a washing solution.

The concentrator may additionally be equipped with a pressure compensation mechanism.

The solution comprising larger sized radioactive compounds which is to be purified can initially be in the form of a solution, a dry powder which must be reconstituted before purification or be in a frozen form which must thaw before purification.

The (post) washing solution will typically be a sterile and pyrogen-free buffer solution, bacteriostatic water, glucose solution or isotonic saline for injection all of which are acceptable for administration by injection to a patient. The washing solution may additionally contain proteins, if desirable.

The kit can additionally contain one or more sterile syringe(s) or sterile container(s), preferably a sterile syringe having suitable volume(s). If the concentrator is not provided in a pre-equilibrated condition, the kit can further encompass a pre-wash solution comprising proteins as described above to reduce absorption. If appropriate, the kit can be provided with analytical tools such as for example thin-layer chromatography (TLC) plate(s) to carry out analysis that secure that the desired purity of the product has been reached.

The kit is furthermore contemplated to encompass a means for shielding of the concentrator during purification if necessary. An appropriate means for shielding is by way of example a lead shield container which reduces radiation exposure to the operator. The container should preferably be provided with an opening, aperture or window allowing the operator follow the progress of the purification process. Optionally the whole kit can be packaged in a shielding material. The kit can also contain a packaging enclosure, e.g. describing the contained parts of the kit, the use of the kit and other essential information normally given in such packaging enclosures in the pharmaceutical field.

In one embodiment the kit contains:

5-50 ml of radioimmunoconjugate solution in a sterile syringe, 5-50 ml of bacteriostatic water/isotonic saline for injection in a sterile syringe, and the disposable, sterile, pyrogen-free and ready-to-use concentrator of the present invention.

In another embodiment the kit contains: 5-50 ml of 5 % human serum albumin,

5-50 ml of radioimmunoconjugate solution in a sterile syringe, 5-50 ml of bacteriostatic water/isotonic saline for injection in a sterile syringe, and the disposable, sterile and pyrogen-free concentrator of the present invention.

All the parts of the kit having been in contact with the radioactive solution should preferentially be disposed as radioactive material after use of the kit.

A further aspect of the invention relates to a method for purifying a solution comprising larger sized radioactive compounds comprising the following steps

- injecting a solution comprising larger sized radioactive compounds into the concentration chamber of the pre-washed concentrator; - waiting for a predetermined time < 3 hours;

- withdrawing the concentrated solution comprising larger sized radioactive compounds from the concentration chamber into a syringe containing washing solution;

- withdrawing the whole volume from the concentration chamber into the same syringe again.

The steps of this method are illustrated in Figure 2. Initially the outer cap protecting the rubber-cap is partly or totally removed and if the sterile and pyrogen-free concentrator is not pre-equilibrated, it should be washed with the pre-wash solution described above before use. It might be necessary to empty the concentrator if it contains a pre-wash solution introduced by the manufacturer. When the concentrator is ready for use it is placed in a shield, if necessary, and the solution comprising larger sized radioactive compounds is introduced into the concentrator by injection trough the rubber-cap using a sterile syringe. If the radioactive product to be purified is in the form of a powder it has to be reconstituted by addition of a liquid being acceptable for injection or in the case the radioactive product is frozen it must be thaw. After injection of the radioactive solution, the operator has to wait until the solution has been concentrated to a certain level or concentration optionally for a predetermined time. The waiting period required will be from 15 to 180 minutes, preferably 15 to 120 minutes, e.g. 20 to 90 minutes, 30 to 70 minutes, 40 to 65 minutes and typically about 1 hour. By this time the initial volume has been concentrated to 1-20% of the initial volume, while simultaneously 99-80% of the liquid and small molecular compounds have been removed. Then the concentrated solution comprising larger sized radioactive compounds is withdrawn from the concentration chamber into a syringe containing washing solution suitable for administration to dilute the radioactive product. A volume, typically of 70-100% of 1-50 ml, of the diluted solution comprising larger sized radioactive compounds is injected into the concentration chamber of concentrator again to wash out the remaining parts of the product. Finally the whole volume from the concentration chamber is withdrawn into the syringe again. The finally purified sterile solution comprising larger sized radioactive compounds withdrawn into the syringe can be quickly analyzed by an appropriate technique and then administered to the patient by injection.

In an alternative embodiment the invention provides a method for purifying a solution comprising larger sized radioactive compounds substantially as described above but wherein the concentration chamber is washed after removal of the concentrated radioactive compounds by use of a second syringe with washing solution. Thus, after the removal of the purified concentrated solution comprising larger sized radioactive compounds from the concentration chamber into a first syringe containing washing solution, a second syringe is used for injecting a volume of washing solution into the concentration chamber of concentrator, and this volume is withdrawn from the concentration chamber into the second syringe again. The content of both the first and the second syringes are used for injection into the patient.

Still another aspect of the invention concerns use of a kit as disclosed above to prepare a sterile and pyrogen-free, preferably sterile solution comprising larger sized radioactive compounds having a radiochemical purity of at least 70%, preferably 95%, more preferred 97%, even more preferred 99% or higher.

Examples

Example 1

The following materials were used in the experiments: Radioimmunoconjugate in solution : 5 ml Of ²²⁷Th-IgG conjugate dissolved in PBS.

Concentrator: Sterile plastic Vivapore 5 concentrator with a 7500 MWCO polyethersulfone membrane (Sartorius Stedim Biotech, Aubagne, France).

Measurements of radioactivity

The radioactivity of the radioimmunoconjugate solutions were measured before and after concentration by a gamma spectrometer with 2.3 keV (FWHM) resolution at 1.33 MeV and 1.5 keV (FWHM) resolution at 122 keV, nominal volume of 300 cm³ and 60 % relative efficiency (Model GCW6021, Canberra Industries Inc., Oak Ridge, USA). Radioactivity was calculated using the Apex program (Laboratory Suite, Canberra Industries Inc., Oak Ridge, USA). The amount Of ²²⁷Th was calculated using the 236 keV (11.2 % yield) and 256 keV (6.8 % yield) γ-ray lines. The amount Of ²²³Ra was calculated using the 154 keV (5.6 % yield), 269 keV (13,6 % yield), 324 keV (3.9 % yield) and 338 keV (2.8 % yield) γ-ray lines. Results

Experiment 1 25 X concentration of conjugate

25 x concentration of conjugate was attempted. The concentrator was pre-washed with 5 ml of PBS with 4 % Fetal Calf Serum (FCS) and 1 mg/ml cold, i.e. unlabeled, antibody. 5 ml of the ²²⁷Th-IgG solution was added to the 5 ml concentrator. The recovery of ²²⁷Th-IgG was 80 % and the recovery of ²²³Ra was 5 %. The concentration period lasted for 1 hour. Other daughter nuclides were undetectable.

Experiment 2 10 X concentration and same concentrator

Two times 10 X concentration using the same concentrator twice were tested. The concentrator was pre-washed with 5 ml PBS with 4 % FCS and 1 mg/ml cold antibody. In the first concentration process the recovery of ²²⁷Th-IgG was 83 % and the recovery of ²²³Ra was 10 % after 40 min. The concentrated solution was then diluted 10 times in PBS and the solution was again concentrated 10 X. Now the concentration procedure lasted for 2 h. The recovery Of ²²⁷Th-IgG was 81 % and the recovery of ²²³Ra was 4 %. Other daughter nuclides were undetectable.

Experiment 3 Dilution of radioimmunoconjugate solution before concentration

This experiment was performed to test whether dilution before extraction could increase the recovery of ²²⁷Th-IgG. The 5 ml concentrator was washed with 5 ml PBS with 4 % FCS and 1 mg/ml cold antibody. The conjugate solution was concentrated from 5 ml to 0.1 ml (50 X) and was extracted by first injecting 2 ml PBS with 4 % FCS into the concentrator and then extracting the diluted conjugate solution. Recovery of ²²⁷Th-IgG was 77 %, while recovery of ²²³Ra was 4 %. Other daughter nuclides were undetectable. Thus, dilution before extraction did not increase recovery of ²²⁷Th-IgG.

Experiment 4 Different pre-washing solutions and post washing of concentrator

This experiment was performed to find out if pre-washing of the concentrator with PBS containing only 4 % FCS was equal to pre-washing using PBS containing 4 % FCS and 1 mg/ml IgG. Additionally it was desirable to know if washing of the concentrator after extraction of the concentrated conjugate solution resulted in better recovery. Accordingly, the concentrator was pre-washed with 5 ml PBS containing 4% FCS (i.e. without 1 mg/ml IgG) before 25 X concentration. The 5 conjugate 5 ml was extracted, the concentrator was washed with 1 ml PBS solution and this wash was added to the concentrated conjugate. Now 83 % recovery of ²²⁷Th-IgG and 5 % recovery of ²²³Ra were obtained. Thus, the recovery of ²²⁷Th-IgG was slightly higher than without washing and it seems to be unnecessary to include 1 mg/ml IgG in the pre-wash solution. 10

Conclusion

The average recovery of ²²⁷Th-IgG in experiments 1 to 4 was 81 ± 2 %, while the average recovery of ²²³Ra was 6 ± 2 % (Figure 3). The recovery of ²²⁷Th-IgG was highest when the concentrator was washed after concentration of the conjugate. 15 The recovery Of ²²³Ra was lower for 25 X concentration than for 10 X concentration.

EXAMPLE 2

A freshly prepared Yttrium-90 labeled monoclonal antibody ⁹⁰Y-tiuexetan- 20 ibritumomab (Zevalin) was analysed by thin layer chromatography and found to have a radiochemical purity of 89%. It is assumed that 89% of the radionuclide is bound to the antibody and 11% is present as unconjugated radionuclide. According to the protocol, product with a purity < 95% must be discarded. A purification according to Example 1, experiment 1, would yield 80% recovery of 25 ⁹⁰Y-labeled antibody with 0.05x 11% left of the non-conjugated ⁹⁰Y. Thus the purity would be 100% x [0.8x0.89]/([0.8x0.89] + [0.11x0.05]) = 99.2 % radiochemical purity and the batch could therefore be used for clinical application.

EXAMPLE 3

30 Purifying unconjugated iodine-125 from antibody by concentrator

To a 5 ml test tube was added 1 ml Rituximab (10 mg/ml; Mabthera, Roche, Basel, Switzerland) and 10 μl of sodium [¹²⁵I] iodide and the solution was mixed gently by swirling the test tube for 2 min. The test tube was measured for radioactivity on a gamma counter (Cobra Auto-Gamma, Packard, Downers Grove,

35 IL). The activity level was found to be 6.16 MBq. The content was thereafter transferred to a Vivapore 5 concentrator unit (Sartorius Stedim Biotech S. A., Aubagne Cedex, France) and added 4 ml of Dulbecco's phosphate buffered saline (PBS) (Sigma-Aldrich Corporation St. Louis, MO). After 45 minutes the sample had been concentrated from 5 ml to 0.6 ml. The concentrate was withdrawn from the concentration unit and transferred to a 5 ml test tube. In addition the concentrator was washed with 1 ml PBS and the washing solution pooled with the concentrate in the test tube. Thereafter the test tube was counted on a gamma counter. In the concentrate the activity level was 0.36 MBq. Thus 94% of the activity was removed by the concentration method. Conclusion

Anionic radioiodine is effectively removed from monoclonal antibody by the present concentration method.

EXAMPLE 4 Purifying unconjugated indium-Ill from antibody labeled with iodine-125. A test tube containing a mix of 1 ml of ¹²⁵I-labelled antibody (42 μg/ml; HHl-antibody, Diatec, Oslo, Norway) and 0,2 μl of ¹¹¹In as indium-chloride was measured for radioactivity on a gamma counter (Cobra Auto-Gamma, Packard, Downers Grove, IL). The test tube contained 37.1 kBq of ¹²⁵I-HHl and 56.7 kBq of ¹¹¹In. Four ml Dulbecco's phosphate buffered saline (PBS) (Sigma-Aldrich Corporation St. Louis, MO) with 4 % Foetal Calf Serum (FCS) was added to the test tube. The content was thereafter transferred to a Vivapore 5 concentrator unit (Sartorius Stedim Biotech S. A., Aubagne Cedex, France) that was previously washed with PBS with 4 % FCS. After 60 minutes the sample had been concentrated from 5 ml to 0.5 ml. The concentrate was withdrawn from the concentration unit and transferred to a 5 ml test tube. In addition the concentrator was washed with 1 ml PBS and the washing solution pooled with the concentrate in the test tube. Thereafter the test tube was counted on a gamma counter. In the concentrate the activity of ¹²⁵I-HHl was 34.4 kBq and the activity of ¹¹¹In was 2.6 kBq. Thus 95% of the ¹¹¹In activity was removed by the concentration method and 93 % of the radiolabeled antibody was recovered. Conclusion

Unconjugated ¹¹¹In was effectively removed from radioimmunoconjugate using the Vivapore 5 unit indicating that the method is relevant for purifying radioimmunoconjugates labeled with indium or similar cations. EXAMPLE 5 - Sephadex C25 filtration - a comparative purification method

A 0.4 ml solution containing ²²⁷Th-IgG and free ²²³Ra and daughter nuclides was, after counting by a digital gamma ray spectrometer (GCW6021, Canberra. Meriden, CT), filtered using a column made of 0.8 ml of Sephadex C25 gel

(Aldrich, St. Louis, MI) placed in a 2 ml syringe. The sample was loaded onto the column and eluted using PBS. Five fractions of 0.5 ml each were collected and counted with the gamma spectrometer. The recovery of ²²⁷Th-IgG in the collected fractions was 65 % and no ²²³Ra was detected in any of the collected fractions, i.e., at least 95% of the radium was retained on the column. Conclusion

The sephadex C-25 material was effectively removing the ²²³Ra cations from the radioimmunoconjugate but unfortunately more than 1/3 of the radioimmunoconjugate was lost during the procedure, suggesting a significant absorption of protein to the ion exchange material.

EXAMPLE 6 - Traditional dialysis - a comparative purification method

Dialysis of a solution containing radioimmunoconjugate and unconjugated radionuclide A solution containing ²²⁷Th-IgG and the daughter nuclide ²²³Ra was, after counting by a digital gamma ray spectrometer (GCW6021, Canberra. Meriden, CT), placed inside a dialysis tube {.Mini Dialysis kit, GE Healthcare Bio-Sciences, Uppsala, Sweden). The dialysis tube was positioned in a beaker containing (1) PBS or (2) sterile water with volumes at least 10 times that of the sample in the dialysis tube.

(1) After 16 h of dialysis of 2 ml ²²⁷Th-labeled antibody against 50 ml of PBS, 85% of the ²²⁷Th and 67% the ²²³Ra was associated with the solution inside the dialysis tube.

(2) After 4 h 15 min of dialysis of 1 ml ²²⁷Th-labeled antibody against 50 ml sterile water 90.5% of the ²²⁷Th and 43.5 % of the initial ²²³Ra was associated with the large molecular bound fraction inside the dialysis tube. The experiment was repeated with 0.5 ml ²²⁷Th-labeled antibody against 50 ml of distilled water. After 18 hours 97% of the ²²⁷Th-labeled antibody and 45% of the ²²³Ra was associated with the solution inside the tube.

Finally an experiment using 0.5 ml internal and 500 ml external solution (distilled 5 water) and also magnetic stirring of the external solution for 48 h gave only 50% purification Of ²²³Ra from the ²²⁷Th-labeled antibody solution.

Conclusion

Although dialyses were tested over several hours, a sufficient purification of free 10 ²²³Ra from that of ²²⁷Th-labeled antibody was not achieved.

EXAMPLE 7 Purifying ¹²⁵I-labeled antibody using a centrifuge based filtration unit - a comparative purification method

15 A 10 ml (0.1 mg/ml) sample of radioimmunoconjugate was added to the centrifugal concentration cartridge (UFV2BTK10, 30KNMWL membranes, 15 ml maximum volumes, Millipore, Bedford, IL, USA). Together with a balance weight the cartridge was placed in a centrifuge (Centrifuge 5810R, Eppendorf, Germany) operated at 20 ⁰C and 1 hour of centrifugation on 1400 rcf. After 1 hour of

20 centrifugation the sample was concentrated to 0.7 ml. The recovery of the sample was only 68% indicating that absorption of protein was a problem at least when low protein concentrations were used.

EXAMPLE 8

25 A sterile version of the Vivapore 5 concentrator unit (Sartorius Stedim Biotech S. A., Aubagne Cedex, France) was made by plugging the hole for inserting solution with a rubber cone. Furthermore, a second hole was drilled on top of the concentrator on the other side of the membrane, i.e., on top of the absorption chamber, which do not have to be sterile. This hole was also plugged with a

30 rubber cone and a sterile needle was pinned through it. This second hole was necessary in order to let air escape from the concentrator when solution is injected through the rubber cone of the first hole. Subsequently, the concentrator was autoclaved at 60 ⁰C for 2 hours. In order to test the sterility of the concentrator, 5 ml cell growth medium was

35 injected with a syringe via a needle pinned through the rubber cone. After 5 minutes the growth medium was drawn from the concentrator and injected into a cell growth flask, which was placed in an incubator with a humidified atmosphere of 5 % CO₂ and temperature of 37 ⁰C. Every day for 7 days the flask was checked for growth of bacteria by a light microscope. No bacteria were detected in the flask.

Conclusion

This example shows that the technology can be used for sterile purification.

Thus, the purified solution comprising the larger sized radioactive compounds purified by the sterile and pyrogen-free concentrator of the present invention will have the necessary radiochemical purity and sterility making the solution acceptable for administration by injection directly into a subject, such as an animal or human being.

Although the present invention has been described in connection with the specified embodiments, it should not be construed in any way as being limited to the presented examples. The scope of the present invention is to be interpreted in the light of the accompanying claim set. The use of reference signs in the claims with respect to elements indicated in the figures shall not be construed as limiting the scope of the invention.

Claims

PATENT CLAIMS:

1. Kit for purification of a solution comprising larger sized radioactive compounds, which solution can be administrated directly into a patient, comprising

- a concentrator for removing or replacing unconjugated radionuclides in a solution comprising larger sized radioactive compounds without applying centrifugation, pressure or vacuum, comprising a concentration chamber for receiving the solution, an absorption chamber for receiving at least the unconjugated radionuclides provided with absorbent material, the concentration and the absorption chambers being connected through a permeable membrane, where the concentrator is a sterile and pyrogen-free closed unit having an opening for transferring and withdrawing solution sealed with a syringe-penetrable cap;

- a solution containing larger sized radioactive compounds; and

- a washing solution.

2. The kit according to claim 1, wherein the concentrator is equipped with a pressure compensation mechanism.

3. The kit according to claim 1, wherein the larger sized compounds are radiolabeled proteins, liposomes or nano-particles, preferably ⁹⁰Y-tiuexetan- ibritumomab (Zevalin™), ¹³¹I-tositumomab (Bexxar™), ²²⁷Th-labeled and ²²⁵Ac- labeled targeting molecules including antibodies and antibody fragments.

4. The kit according to claim 1, wherein the washing solution is selected from a buffer solution, bacteriostatic water, glucose solution or isotonic saline for injection.

5. The kit according to claim 1, wherein the kit further comprises a pre-wash solution, preferably saline with 5% HSA or PBS with 4% FCS.

6. The kit according to claim 1, wherein the kit further comprises a lead shielding container.

7. Method for purifying a solution comprising larger sized radioactive compounds, which solution can be administrated directly into a patient, the method comprising of the following steps

- injecting a solution comprising larger sized radioactive compounds into the concentration chamber of a sterile pre-washed concentrator according to claim 1 or 2;

- waiting for a predetermined time < 3 hours;

8. The method according to claim 7, wherein the predetermined waiting period is from 15 to 120 minutes, preferably about 1 hour.

9. Use of a kit according to claim 1 or 2 to prepare a sterile solution comprising larger sized radioactive compounds having a radiochemical purity of at least 70%, more preferably at least 95%.

10. Concentrator for removing or replacing unconjugated radionuclides in a solution comprising larger sized radioactive compounds without applying centrifugation, pressure or vacuum, comprising a concentration chamber for receiving the solution, an absorption chamber for receiving at least the unconjugated radionuclides provided with absorbent material, the concentration and the absorption chambers being connected through a permeable membrane, the concentrator being

- a sterile and pyrogen-free closed unit having

- an opening for transferring and withdrawing solution sealed with a syringe- penetrable cap; and being

- equipped with a pressure compensation mechanism.

11. The concentrator according to claim 10, wherein the pressure compensation mechanism is a capillary.

12. The concentrator according to claim 10, wherein the pressure compensation mechanism is a syringe needle penetrating the syringe-penetrable cap.

13. The concentrator according to claim 10, wherein the pressure compensation mechanism is a two-way valve that allows flow of fluid only when the pressure inside the concentrator is higher or lower that the surrounding pressure.

14. The concentrator according to any of claims 10-13, wherein the pressure compensation mechanism is equipped with a sterile filter.