Automated processing of solid target 86Y using enriched SrO powder.

86Y (t1/2 = 14.74 h, 32% ß+) has significant potential in theranostic applications as a simultaneous PET imaging partner to 90Y-labelled antibody therapy. However, the complex and costly nature of producing 86Y has led to this radiometal being difficult for hospitals and researchers to obtain. The aim of this work was to develop a simple and cost-efficient method for safely producing 86Y. Our approach was twofold: to develop a method of target preparation that would significantly increase the cost efficiency of producing 86Y, and to design and construct an automated purification system that would eliminate manual radiation handling risks and exposure. Multiple automated productions of high radionuclidic purity (99.45%) 86Y were performed resulting in saturation yields of between 518 MBq/µA and 1332 MBq/µA, dependent on target thickness.

Facile radiochemical separation of clinical-grade 90Y from 90Sr by selective precipitation for targeted radionuclide therapy.

INTRODUCTION: The widespread clinical utilization of 90Y for preparation of target specific radiopharmaceuticals demands development of a facile, efficient and cost-effective method for radiochemical separation of 90Y from 90Sr via90Sr/90Y generator. In this article, we describe an efficient and facile method for radiochemical separation of 90Y from 90Sr for preparation of radiopharmaceuticals by exploiting the large difference in the solubility product constants (Ksp) of Y(OH)3 and Sr(OH)2. METHODS: A two-step radiochemical separation procedure based on selective precipitation of 90Y under alkaline conditions from 90Sr/90Y equilibrium mixture was developed. The 90Y(OH)3 colloid formed at pH ~ 10 was selectively trapped in 0.22 µm sterile filter and was subsequently retrieved by dissolution in HCl solution. Detailed quality control analyses of obtained 90Y were carried out and its utility towards preparation of different radiopharmaceuticals was assessed. RESULTS: Using the same feed solution of 90Sr (3.7 GBq), consistent and repeated separation of 90Y could be achieved in different batches with >85% yield and >99.999% radionuclidic purity. Yttrium-90 obtained from this process was found suitable for preparation of therapeutically relevant doses of three different radiopharmaceutical formulations, namely, 90Y-DOTA-TATE, 90Y-PSMA-617 and 90Y-CHX-A″-DTPA-Cetuximab with >95% radiochemical purity. CONCLUSIONS: The promising results obtained in this study would facilitate implementation of the developed technique for obtaining 90Y in adequate quantity and of required purity from a centralized radiopharmacy setup.

A preliminary study for the production of high specific activity radionuclides for nuclear medicine obtained with the isotope separation on line technique.

Radiopharmaceuticals represent a fundamental tool for nuclear medicine procedures, both for diagnostic and therapeutic purposes. The present work aims to explore the Isotope Separation On-Line (ISOL) technique for the production of carrier-free radionuclides for nuclear medicine at SPES, a nuclear physics facility under construction at INFN-LNL. Stable ion beams of strontium, yttrium and iodine were produced using the SPES test bench (Front-End) to simulate the production of 89Sr, 90Y, 125I and 131I and collected with good efficiency on suitable targets.

Analysis of radioactive strontium-90 in food by Cerenkov liquid scintillation counting.

A simple liquid scintillation counting method using DGA/TRU resins for removal of matrix/radiometric interferences, Cerenkov counting for measuring 90Y, and EDXRF for quantifying Y recovery was validated for analyzing 90Sr in various foods. Analysis of samples containing energetic ß emitters required using TRU resin to avoid false detection and positive bias. Additional 34% increase in Y recovery was obtained by stirring the resin while eluting Y with H2C2O4. The method showed acceptable accuracy (±10%), precision (10%), and detectability (~0.09Bqkg-1).

Formulation and purification of therapeutic dose of 90Y-labeled peptides: Some interesting radiochemistry aspects.

Yttrium-90 obtained from most of the 90Sr/90Y generators contains 90Sr impurity above permissible limit for human administration. A protocol has been optimized for formulation of therapeutic dose of 90Y-DOTA-Tyr3-octreotate (90Y-DOTA-TATE) and removal of 90Sr impurity from it. The radiochemical purity of 90Y-DOTA-TATE was found to be >98% and it met the requirements for clinical use. The radiopharmaceutical was used in preliminary clinical investigation in patients with neuroendocrine tumors. This promising strategy would aid toward widespread clinical utilization of 90Sr/90Y generators.

Production of Y-86 and other radiometals for research purposes using a solution target system.

INTRODUCTION: Diagnostic radiometals are typically obtained from cyclotrons by irradiating solid targets or from radioisotope generators. These methods have the advantage of high production yields, but require additional solid target handling infrastructure that is not readily available to many cyclotron facilities. Herein, we provide an overview of our results regarding the production of various positron-emitting radiometals using a liquid target system installed on a 13 MeV cyclotron at TRIUMF. Details about the production, purification and quality control of (89)Zr, (68)Ga and for the first time (86)Y are discussed. METHODS: Aqueous solutions containing 1.35-1.65 g/mL of natural-abundance zinc nitrate, yttrium nitrate, and strontium nitrate were irradiated on a 13 MeV cyclotron using a standard liquid target. Different target body and foil materials were investigated for corrosion. Production yields were calculated using theoretical cross-sections from the EMPIRE code and compared with experimental results. The radioisotopes were extracted from irradiated target material using solid phase extraction methods adapted from previously reported methods, and used for radiolabelling experiments. RESULTS: We demonstrated production quantities that are sufficient for chemical and biological studies for three separate radiometals, (89)Zr (Asat = 360 MBq/µA and yield = 3.17 MBq/µA), (86)Y (Asat = 31 MBq/µA and yield = 1.44 MBq/µA), and (68)Ga (Asat = 141 MBq/µA and yield = 64 MBq/µA) from one hour long irradiations on a typical medical cyclotron. (68)Ga yields were sufficient for potential clinical applications. In order to avoid corrosion of the target body and target foil, nitrate solutions were chosen as well as niobium as target-body material. An automatic loading system enabled up to three production runs per day. The separation efficiency ranged from 82 to 99%. Subsequently, (68)Ga and (86)Y were successfully used to radiolabel DOTA-based chelators while deferoxamine was used to coordinate (89)Zr.

Production and quality control of radioactive yttrium microspheres for medical applications.

In this paper, a method for production of yttrium silicate microspheres is reported. Yttrium silicate microspheres with approximate sizes of 20-50µm were obtained when an aqueous solution of Y(NO3)3 was added to tetraethyl orthosilicate (TEOS) and was pumped into silicone oil under constant stirring. The shapes of the particles produced by the proposed method were regular and nearly spherical. The spherical shapes, composition and element distribution were investigated by scanning electron microscopy (SEM), carbon/sulfur analysis and SEM/EDS mapping analysis. Paper chromatography was used to identify radiochemical impurities in the radioactive microspheres. The radionuclide purity was determined using a gamma spectrometry system and an ultra-low-level liquid scintillation spectrometer. The results indicated that the proposed silicone oil spheroidization method is suitable for the production of yttrium silicate microspheres.

A novel approach for preferential recovery of Sr from (Sr, Th)O2.

Quantitative leaching of Sr from homogeneous and calcined (Th,Sr) O(2) in dilute perchloric acid medium suggests the possibility of reducing the hazardousness of discharged nuclear fuel by separation of (90)Sr, a prominent fission product at dissolution stage itself rather than the conventional approach of its recovery from high level nuclear waste. Apart from mitigating the radiotoxicity of the nuclear waste, recovered (90)Sr can be employed as a compact heat source and as parent radionuclide for (90)Y (used in therapy radiopharmaceuticals), provided it can be made available at desired high purity. Leaching behavior of few other fission products was also investigated to quantify their contamination in leached Sr. Feasibility of employing extraction chromatography using Sr selective resin was explored in perchloric acid medium. In this context, the distribution coefficients of (85)Sr(II), Th (IV), Zr(IV), Y(III), Pd(II) as well as (152)Eu(III) and (137)Cs (I) were determined under varying nitric acid/perchloric acid concentration and under varying loading conditions of metal ions. Perchloric acid medium appears better than nitric acid medium for preferential leaching of Sr from (Th,Sr)O(2) as well as for uptake of Sr by Sr selective chromatographic resin.

Yttrium-90 -- current status, expected availability and applications of a high beta energy emitter.

Yttrium-90 ((90)Y, T(1/2) 64.14 h) is a key example of a high beta energy-emitting radionuclide which is available from the strontium-90 ((90)Sr)/(90)Y radionuclide generator system. Clinical uses of (90)Y-labeled radiopharmaceutical agents have been pursued for many years and many applications have proven to be clinical effective. These most notably include the application of 90Y-labeled antibodies for a variety of applications such as for effective treatment of non-Hodgkin's lymphoma. One of the major advantages for use of (90)Y is ready availability from the very long-lived (90)Sr parent (T(1/2) 28.78 y). Because of the importance of maintaining generator performance and minimizing parent breakthrough, this paper describes development, use and quality control of both high capacity cation adsorption-type and electrochemical generator systems. In addition, the preparation and targeting to tumors in mice of DOTA-conjugated Nimotuzamab (h-R3) antibody which recognizes the external domain of the EPFR antibody radiolabeled with (90)Y obtained from the electrochemical generator is also described. As a key example for clinical applications of (90)Y, the use of (90)Y-labeled biotin for intra-operative pre-targeting for radionuclide therapy (IART®) of breast cancer is also described.

Chromatographic separation of carrier free 90Y from 90Sr using a diglycolamide based resin for possible pharmaceutical applications.

N,N,N',N'-tetraoctyl diglycolamide (TODGA) has been used as the stationary phase in an extraction chromatography resin (XCR) material prepared for evaluating the uptake and the separation behaviour of (90)Y and (90)Sr from acidic feeds. Chromosorb-W was used as the solid support material while the feed solution was usually 4M HNO(3). The batch uptake studies have suggested almost no Sr(II) uptake while Y(III) uptake increased with acidity up to 4M HNO(3) beyond which a decrease in the K(d,w) values were observed. Column studies were carried out and breakthrough profiles were obtained for both Y(III) and Sr(II). No breakthrough of Y(III) was noticed even when >50 column volumes of the feed (carrier free (90)Y at 4 M HNO(3)) was passed through the column while about 20 column volumes were required for the breakthrough of Y(III) when the feed contained 1 g/L Y in 4 M HNO(3) spiked with (90)Y tracer. The reusability of the column was also studied which indicated in the deterioration of the column performance as shown by the sharp fall in the breakthrough volumes and was attributed to the probable leaching of the reagent from the support material. The role of absorbed dose was also investigated for Y(III) uptake. Separation of carrier free (90)Y tracer was carried out by loading the column with (90)Sr and eluting with 0.01M solutions of HNO(3) as well as EDTA. The purity of the product was ascertained by half-life method.

Separation of 90Y from 90Sr by a solvent extraction method using N,N,N',N'-tetraoctyl diglycolamide (TODGA) as the extractant.

The relative extraction efficiency of N,N,N',N'-tetraoctyl diglycolamide (TODGA) towards Y(3+) and Sr(2+) was investigated from HNO(3) as well as HCl medium. The separation factor (S.F.) values were found to be much higher in HCl medium as compared to those obtained from HNO(3) medium. Based on the high separation factor values obtained in the HCl media, a separation method for (90)Y from (90)Sr (containing (90)Y in secular equilibrium) has been developed in the present work. The purity of the separated (90)Y was ascertained from its decay profile and half-life measurements. The extracted species was determined to be Y(X)(3) · 3TODGA where X(-) is the nitrate or the chloride anion. The thermodynamic parameters were also determined and the two phase extraction constants (log K(ex)) were calculated. The extraction of Y(III) was highly exothermic and entropy destabilized and was equally favourable from both HNO(3) medium as well as HCl medium.

Production of large quantities of 90Y by ion-exchange chromatography using an organic resin and a chelating agent.

The performance of a system composed of an organic cation exchanger (Dowex 50Wx8) and a chelating agent (EDTA) previously described for the successful production of (90)Y via a (90)Sr/(90)Y generator is assessed under dynamic conditions. In an attempt to overcome the established limitation of ion-exchange resins for the separation of subcurie quantities of activity, (90)Y is repeatedly isolated from an 11.8-GBq (320 mCi) (90)Sr cow using a three-column tandem arrangement. The high recovery and radionuclidic purity obtained for (90)Y and the parameters of the separation (time, eluant concentration, pH and flow rate range) strongly suggest that Ci quantities of (90)Y can be handled satisfactorily by the ion-exchange method. No replacement or treatment of the cow, low waste generation and (90)Sr losses less than 0.1% after each run were observed during the present study which, in combination with the low cost of this resin, may result in an attractive alternate method for the production of large quantities of (90)Y.

One column operation for (90)Sr/(90)Y separation by using a functionalized-silica.

Organo-ceramic hybrid materials have been developed as the separation media for a (90)Sr/(90)Y generator system. Currently available (90)Y is generally extracted from a mother solution by a solvent extraction or a successive column operation. Both processes are successfully applied to produce (90)Y with a high quality standard. However, such processes are highly dependent on what kind of extracting materials are employed. Hence, some of the previously developed technologies are not adequate for a (90)Y production because of a waste generation or leaching of radiolytic end-products from the extracting materials. In this study, high performance organo-ceramic hybrid materials have been applied for the extraction of (90)Y. The hybrid materials have properties of both a ceramic and a solvent extractant by molecularly implanting the extracting molecules on to the ceramic surfaces. In this study, organo-phosphorus functionalized hybrid materials are synthesized and tested as the separation media for the (90)Y/(90)Sr generator system. An adsorptive extraction with a small Sep-Pak type column can recover more than 92% of (90)Y with a contamination ratio of (90)Sr/(90)Y=1.2x10(-5) from the mother solution and 70% with 5x10(-7).

(86)Y production via (86)Sr(p,n) for PET imaging at a cyclotron.

Excitation functions of (86)Y production via (86)Sr(p,xn), (86)Sr(d,xn), (85)Rb(alpha,xn), (85)Rb((3)He,xn), and (nat)Zr(d,alphaxn) reactions were studied by means of ALICE-ASH code and the results were compared with ALICE-91 code and experimental data. The greatest nuclear reaction of cyclotron (86)Y production was found out as (86)Sr(p,n)(86)Y process. (86)Y production yield was calculated too. A SrCO(3) thick film was deposited on a copper substrate by sedimentation method. The deposited (nat)SrCO(3) was irradiated with 15MeV proton at 30microA current beam. The separation of Y from Cu and Sr was carried out by means of dual ion exchange chromatography.

Radiochemical studies relevant to 86Y production via 86Sr(p,n)86Y for PET imaging.

A novel production technique of yttrium-86 based on bombardment of deposited strontium carbonate was investigated. (86)Y was produced via proton-induced reactions on SrCO(3) target that was prepared by the sedimentation method. Production yield of 0.37mCi/microAh at 30 microA was measured by means of gamma-ray spectrometry for natural target. The separation of (86/87/88)Y from Cu and Sr was carried out by two ion-exchange columns.

Production and separation of non-carrier-added 86Y from enriched 86Sr targets.

The metallic radionuclide (86)Y was produced by irradiation of enriched (86)SrCO(3) on a low-energy proton-only cyclotron. Irradiations up to 20 microA for 2h were performed with 11 MeV protons using a water-cooled target mounting with circulating chilled helium. Experimental thick target yields of 26.7 mCi/microA yielded 24 mCi of (86)Y in 2h of bombardment at 10 microA. The difference in solubility products between Y(OH)(3) and Sr(OH)(2) allows the separation of (86)Y from an alkaline strontium solution by using filter paper with an overall yield of 88 +/- 3%. The concentration of Sr in the final product was found to be on the order of 15 ppm when using 200mg of target material as determined by ICP-MS analysis. The reactivity of (86)Y was determined to be on the order of 1.5 +/- 0.8 Ci/micromol of DOTA. The enriched target material was recovered and converted to its original chemical form with an overall efficiency >90%.

Semi-automated 86Y purification using a three-column system.

The separation of (86)Y from (86)Sr was optimized by a semi-automated purification system involving the passage of the target sample through three sequential columns. The target material was dissolved in 4 N HNO(3) and loaded onto a Sr-selective (Sr-Spec) column to retain the (86)Sr. The yttrium was eluted with 4 N HNO(3) onto the second Y-selective (RE-Spec) column with quantitative retention. The RE-Spec column was eluted with a stepwise decreasing concentration of HNO(3) to wash out potential metallic impurities to a waste container. The eluate was then pumped onto an Aminex A5 column with 0.1 N HCl and finally with 3 N HCl to collect the radioyttrium in 0.6-0.8 mL with a >80% recovery. This method enabled us to decontaminate Sr by 250,000 times and label 30 micro g of DOTA-Biotin with a >95% yield.

A new and convenient method for purification of 86Y using a Sr(II) selective resin and comparison of biodistribution of 86Y and 111In labeled Herceptin.

A simple and rapid procedure was developed for purification of cyclotron produced 86Y via the 86Sr(p,n) 86Y reaction. A commercially available Sr(II) selective resin was used to separate 86Y from the cyclotron irradiated Sr(II) target with a recovery of the enriched Sr(II) target while yielding a 75-80% recovery of 86Y suitable for radiolabeling either proteins or peptides. To demonstrate the utility of this methodology, the anti-HER2 monoclonal antibody Herceptin was radiolabeled with the purified 86Y and compared to 111In labeled Herceptin. The biodistribution study demonstrated that 111In-Herceptin, while a suitable surrogate for 90Y in the major organs, did not parallel the uptake of 86Y-Herceptin in the bone, and thus may not accurately predict the level of 90Y accumulation in the bone for clinical RIT applications. This result exemplifies the requirement of employing appropriate matched pair isotopes for imaging and therapy to insure that dosimetry considerations may be addressed accurately.

Preparation and studies with 90Y-labelled particles for use in radiation synovectomy.

90Y-FHMA (Ferric hydroxide macroaggregates) and 90Y-HA (hydroxyapatite) were prepared in >95% yield using 90Y from an in-house 90Sr-90Y generator. Most of the particles ranged from 5 to 20 microm in diameter and retained radiochemical purity > 95% in human serum for at least 7 days at 37 degrees C. Gamma camera imaging of normal rabbits after intraarticulation of the particles showed complete retention of activity within the knee cavity with no measurable radioactivity leaching out of the joints over a 96 hour period.

[Determination of 90 Sr in milk by solvent extraction of 90Y (author's transl)].

In order to replace the conventional method using violent fuming nitric acid, a new method for the determination of 90 Sr in milk has been developed by using the solvent extraction with bis (2-ethylhexyl) phosphoric acid (HDEHP). The daughter nuclide 90Y in a radiochemical equilibrium with its parent 90Sr was extracted with 2:1 HDEHP-toluene from the acid solution (1M HCL) of milk ash sample prepared by dry-ashing. After stripping with 8M HCL, 90Y, together with stable yttrium added as carrier, was precipated as oxalate to prepare beta-counting source. The radiochemical purity was confirmed by decay curve. The decontamination of strontium was checked by applying non-dispersive fluorescence X-ray analysis using 133Ba as irradiating source. Bone samples of cow were also analyzed by the same method and the results were compared with those obtained by other methods. The duplicate crosschecking analyses of finely ground bone samples were carried out to examine the effectiveness of this method. This simple new method was found to be very effective for the routine analysis of 90Sr in these samples.