A 228Ra-212Pb tandem generator for potential application in biomedical studies

A new tandem generator system (cation-exchange generator column coupled to anion-exchange), based on the ²²⁸Ra as starting radionuclide, for ²¹²Pb/²¹²Bi production in the solutions suitable to direct application in biomedical investigations has been developed. Optimum conditions have been found for retention of Ra, Th and Ac parent radionuclides on the cation-exchange column and repeated elution of Pb with subsequent concentration on anion-exchange column using HBr and mixed HBr/CH₃OH solutions. It was shown that Pb/Bi could be eluted from anion — exchange column with a small volume of EDTA, DTPA, NaCl solutions or heated H₂O. The advantages of the tandem generator system for production of short-lived Pb/Bi radionuclides and their application for biomedical studies are discussed.     

Rapid separation of212Pb−212Bi−208Tl by high-performance liquid ion chromatography

Fast radiochemical separation of carrier-free²¹²Pb–²¹²Bi–²⁰⁸Tl in radioactive equilibrium was performed using the HPLC technique with an ion-exchange column /ION-210 for cation exchange, ION-110 for anion exchange and C18 loaded with cation-exchange sites/. Optimal results are obtained with the ION-110 column.²⁰⁸Tl and²¹²Pb are eluted by 0.5M HCl in 2.5 and 5 min, respectively.²¹²Bi is recovered by elution with 1M HNO₃.     

Recovery of 177Lu from Irradiated HfO2 Targets for Nuclear Medicine Purposes

A new method of production of one of the most widely used isotopes in nuclear medicine, ¹⁷⁷Lu, with high chemical purity was developed; this method includes irradiation of the HfO₂ target with bremsstrahlung photons. The irradiated target was dissolved in HF and then diluted and placed onto a column filled with LN resin. Quantitative sorption of ¹⁷⁷Lu could be observed during this process. The column later was rinsed with the mixture of 0.1 M HF and 1 M HNO₃ and then 2 M HNO₃ to remove impurities. Quantitative desorption of ¹⁷⁷Lu was achieved by using 6 M HNO₃. The developed method of ¹⁷⁷Lu production ensures high purification of this isotope from macroquantities of hafnium and zirconium and radioactive impurities of carrier-free yttrium. The content of ^177mLu in ¹⁷⁷Lu in photonuclear production was determined. Due to high chemical and radionuclide purity, ¹⁷⁷Lu obtained by the developed method can be used in nuclear medicine.     

Chemical separation of enriched cadmium target from copper backing in cyclotron production of radioisotope 111In

A radiochemical purification procedure was developed for the separation of enriched cadmium (¹¹¹Cd and ¹¹²Cd) from natural copper that used as backing; and was based upon the chromatographic adsorption. The separation of copper from cadmium was studied in this work. The ions were selectively separated from aqueous solution. Ion-exchange chromatography was employed as a column (1.5 cm i.d. and 15 cm length) with AG1-X8 resin (chloride form, 100–200 mesh) and a flow rate of 1–2 ml/min throughout the separation. 6 M HCl media was used for the adsorption of Cd and Cu on the resin. Then, Cu was eluted by 2 M HCl and Cd by 100 ml 0.5 M HNO₃. The amount of Cu and Cd ions in the final solution (0.5 M HNO₃) were measured by pulse polarographic method and the concentration of Cu was found to be <0.1 ppm. The Cd was quantitatively recovered and the recovery yield from ion-exchange chromatography was greater than 96 %.     

Cation-exchange separation of uranium from thorium in nitric acid medium</p> </p>

Summary Sorption behavior of Th and U on cation-exchange resins was investigated from nitric acid medium by both batch and column methods. The cation-exchange studies involved the sorption of UO₂²⁺ and Th⁴⁺ and their cationic complexes onto Dowex 50Wx8 and Dowex 50Wx4 resins (50-100 mesh). The batch data yielded a separation factor (K_d,Th/K_d,U) value of >100 for the cation-exchanger, Dowex 50Wx4 at 1-2M HNO₃. Separation of uranium from thorium was also carried out by column method in nitric acid medium using cation-exchangers, Dowex 50Wx4 as well as Dowex 50Wx8. While uranium elution was possible at 1M HNO₃, Th could be eluted only at higher concentration of nitric acid (>6M). The stripped solution emanating from a mixer settler employing di-2-ethyl hexyl isobutyramide as extractant and feed solution similar to THOREX process comprising 350 mg/l U and 380 mg/l Th in 0.75M HNO₃ was loaded on the column and the decontamination factor value for U in the product was >1000.</p> </p>     

An 227Ac-211Pb generator for test experiments of solution chemistry of element 114

A simple and effective method has been developed for the continuous production of the short-lived isotope Pb as a homolog of element 114 on the principle of a radionuclide generator. The method is based on the initial sorption of ²²⁷Ac or ²²³Ra on a small cation-exchange column, with subsequent “milking” of ²¹¹Pb (36.1 m) by a mixture of HCl/CH₃OH. The optimum conditions for the repeated separation of ²¹¹Pb from radionuclides strongly sorbed by the cation-exchanger (Ac, Th, Ra, Pu, TPE) have been determined. Possibilities of using the ²¹¹Pb generator for test experiments on the solution chemistry of element 114 have been shown. Advantages of aqueous alcohol HCl solutions for the isolation of ²¹¹Pb (element 114) are discussed. This revised version was published online in August 2006 with corrections to the Cover Date.     

Chromatographic isolation of144Ce and144Pr from the wastes of irradiated uranium treatment

A two-step chromatographic technique was elaborated to isolate¹⁴⁴Ce,¹⁴⁴Pr from a solution of uranium fission products in 6M HNO₃. The oxidation to Ce(III) by bromate and selective adsorption of¹⁴⁴Ce(IV) on anion exchange column were used to concentrate and purify¹⁴⁴Ce. Some impurities of uranium,⁹⁵Zr,⁹⁵Nb,¹⁰⁶Ru remain in¹⁴⁴Ce solution after the first step of its isolation. The final purification is achieved by passing the 6M HNO₃ solution of¹⁴⁴Ce(IV) through the HDEHP-coated teflon column. The decontamination factors of¹⁴⁴Ce from main fission products are given. 7.2 mCi of (¹⁴⁴Ce+¹⁴⁴Pr) are recovered from each gram of irradiated uranium trioxide with the yield greater than 99%. An improvement of known generator was carried out to elute a purer¹⁴⁴Pr from maternal¹⁴⁴Ce(IV) adsorbed on the anion exchange column.     

Separation of carrier-free210Bi and UX1 from210Pb and U,respectively, using silica gel as ion exchanger

Radiochemical separations of carrier-free²¹⁰Bi and UX₁ activities from²¹⁰Pb and U, respectively, have been carried out using a silica gel column.²¹⁰Pb was adsorbed in the column as molybdate and²¹⁰Bi passed unadsorbed. Lead activity was next removed with 25 ml of 0.1 M HNO₃. In the case of separation of UX₁, the coloured carbonate complex of U was removed from the silica surface by washing with saturated sodium carbonate solution, keeping UX₁ retained, and finally UX₁ was washed out with 25 ml of conc. HNO₃. Studies of the beta decay of²¹⁰Bi and the γ-spectrum analysis of UX₁ has shown that the separated products in both cases are of high radiochemical purity. The processes in each case took less than one hour and the yield was satisfactory.     

Simultaneous production and separation of no-carrier-added 111In, 109Cd from alpha particle induced silver target

A natural silver foil was bombarded by 30 MeV α-particles which produced ¹¹¹In, ¹⁰⁹Cd and ^106mAg in the target matrix. ¹¹¹In and ¹⁰⁹Cd were separated from the Ag target matrix employing ion-exchange chromatography and liquid–liquid extraction (LLX). In the chromatographic separation, the active solution containing the NCA products were adsorbed in the column containing Dowex 50 and were eluted with HNO₃. Bulk silver and ¹⁰⁹Cd were sequentially eluted with 1 M HNO₃. After complete elution of ¹⁰⁹Cd and the bulk, ¹¹¹In was eluted with 1.5 M HNO₃. In the LLX, the NCA ¹¹¹In was extracted to 1 % HDEHP (di-2(ethylhexyl)phosphoric acid) from 10^?2 M HNO₃ solution, leaving cadmium and bulk silver quantitatively in the aqueous phase. The NCA ¹⁰⁹Cd was separated from the bulk Ag by precipitating Ag as AgCl. NCA ¹¹¹In was stripped back quantitatively from HDEHP phase using 8 M HNO₃.     

Separation of Zr in the rubble waste generated at the Fukushima Daiichi Nuclear Power Station

Separation method of Zr using trans uranium resin (TRU resin) and tetra valent actinide resin (TEVA resin) was developed for the analysis of ⁹³Zr contained in the rubble waste. Zr, Nb, and U were quantitatively extracted on the TRU resin from 3 M HNO₃ and striped with 0.01 M HF, in addition, some part of Mo, Hg, Bi, and Th were also included in the stripping solution. The stripping solution was evaporated to eliminate HNO₃ and the residue was dissolved in 0.1 M HF. Finally, Zr was separated from Nb and Mo with the TEVA resin.     

Preparation of 99Mo/99mTc generator based on alumina 99Mo-molybdate (VI) gel

In this work alumina ⁹⁹Mo-molybdate (VI) gel is evaluated as a column matrix for use in the preparation of small chromatographic column type ^99mTc generator. Alumina molybdate (VI) gel is prepared by dissolving inactive MoO₃ with aluminum foil in 5 M NaOH solution containing ⁹⁹Mo radiotracer. After complete dissolution, 0.5 H₂O₂ was added to the reaction mixture solution and acidified to pH 5.5 with concentrated HNO₃. The formed AlMo precipitate was washed with NaNO₃ solution, dried at 50 °C for 24 h and then packed in the form of a chromatographic column for elution of the generated ^99mTc radionuclide with physiological saline solution (0.9 % NaCl). Greater than 86 % of the generated ^99mTc activity is immediately and reproducibly eluted with passing 10 mL of the saline solution through 2.0 g of alumina ⁹⁹Mo-molybdate column bed at a flow rate of about 1.0 mL/min. The high radiochemical ≥98.6 % TcO₄ ^?, radionuclidic ≥99.90 % ^99mTc and chemical purities of the eluates satisfy the specifications for use in nuclear medicine.     

Search of ligands suitable for 212Pb/212Bi in vivo generators

The short half-life of ²¹²Bi and ²¹³Bi limits the application of these radionuclides in α radionuclide therapy. The labeling of biomolecules with ²¹²Pb (mother nuclide of ²¹²Bi) instead of ²¹²Bi or ²¹³Bi has the advantage of obtaining a conjugate with a half-life of 10.6 h, compared with of 60 min for ²¹²Bi or 46 min for ²¹³Bi. Previous attempts to prepare a potential in vivo generator with ²¹²Pb complexed by the DOTA chelator failed, because about 36 % of Bi was reported to escape as a result of the radioactive decay $^{{212}}{\text{Pb}}{\mathop{\longrightarrow}\limits^{\beta ^{ - }}}{^{212}{\text{Bi}}}$ . Herein, we report studies on the stability of the ²¹²Pb complexes with eight selected polydentate ligands, which demonstrate high affinity for 3+ metal cations. From the ligand studied DOTP and BAPTA show a sufficient ²¹²Pb labeling yields but only ²¹²Pb–DOTP complex is stable in isotonic solution of sodium chloride making this way radioactivity level of released ²¹²Bi is below the limit of detection. It should be emphasized that the DOTP complex is stable only in the case when the concentration of free DOTP exceeds 10^?4 M.     

NBD-Cl as a Post-Column Reagent for Primary and Secondary Amines after Separation by Ion-Exchange Chromatography

In this study 4-chloro-7-nitrobenzo-2-oxa-1,3-diazole (NBD-Cl) is proposed as a post-column derivatization (PCD) reagent for the fluorescence detection of aliphatic primary and secondary amines after HPLC separation. Five primary (methylamine, isoamylamine, 2-phenylethylamine, putrescine, and histamine) and one secondary amine (dimethylamine) were separated isocratically on a cation-exchange column using HNO₃ (5 × 10^?3 mol L^?1) as the mobile phase. Post-column derivatization was based on two steps: 1) the derivatization of amines with NBD-Cl in alkaline medium, and 2) the acidification of the resulted mixture in order to minimize the background signal of the reagent and improve dramatically the sensitivity and determination range. The variables of the post-column reaction (concentration of NBD-Cl, buffer concentration and pH, reaction temperature, concentration of HCl, flow rates of the reagents) were thoroughly investigated. Critical chromatographic parameters such as the concentration of HNO₃, the percentage of organic solvent, and the column temperature were also examined to achieve adequate separation. An internal standard of 1,7-diaminoheptane was used. The developed post-column method provides the ability for a fully automated analysis, low detection limits (LODs 20–100 µg L^?1 with S/N = 3), and it requires less sample preparation. The applicability of the proposed analytical scheme was demonstrated by the determination of histamine (HIS) in tuna fish tissues according to US Food and Drug Administration (FDA) guidelines.     

Sequential separation and determination of uranium and thorium isotopes in soil samples with Microthene-TOPO chromatographic column and alpha-spectrometry

In the method, soil was fused together with Na₂CO₃ and Na₂O₂ at 600 °C, uranium and thorium were leached out with HCl, HNO₃ and HF, and HClO₄ was used to eliminate the residual HF through evaporation. The leaching solution (2 M HNO₃) was passed through a Microthene-TOPO column to adsorb uranium and thorium. Thorium was first eluted with 2 M HCl and electrodeposited in 0.025 M H₂C₂O₄ + 0.15 M HNO₃ on a stainless steel disc. Uranium was eluted with a 0.025 M ammonium oxalate solution and also electrodeposited. Both thorium and uranium isotopes on the discs were measured separately by α-spectrometry.     

Peak parking technique for the simultaneous determination of anions and cations

An ion chromatography method is described for the simultaneous determination of anions (Cl^–, NO₃^–, and SO₄^2–) and cations (Na⁺, NH₄⁺, K⁺, Mg²⁺, and Ca²⁺) using a single pump, a single eluent and a single detector. An anion-exchange column modified with chondroitin sulfate C facilitated the elution of the above three anions using 5 mM tartaric acid as the eluent in isocratic mode, whereas the same eluent facilitated the separation of the above five cations on a commercially-available cation-exchange column. The separation columns were connected in series via two six-port switching valves, so the required cation-exchange or anion-exchange separation could be carried out by selecting the appropriate positions for the switching valves. The separations were completed in 30 min.     

Stability of polyvinylpolypyrrolidone against gamma-ray irradiation in HNO3 media

Polyvinylpolypyrrolidone (PVPP) is a resin with adsorption selectivity for U(VI) in HNO₃ media. The stability of this resin has been examined against gamma-ray irradiation using HNO₃ solutions of various concentrations. Samples of PVPP immersed in HNO₃ solutions up to 6 mol·dm^?3 (=M) were irradiated by a ⁶⁰Co source at 3.6 kGy·h^?1 to reach 0.90 MGy at room temperature under ambient atmosphere. As a result, no clear decrease in the capacity for U(VI) when added in 3 M HNO₃ was observed for any samples, or rather the capacity was found to increase by approximately 50 % for PVPP irradiated in 6 M HNO₃. It was indicated from some spectroscopic analyses of the irradiated samples that the irradiation causes the cleavage of the pyrrolidone ring of PVPP. This occurs by the addition of an oxygen atom originating from HNO₃, followed by the formation of straight-chain groups with multiple coordinative atoms by the successive addition of oxygen, finally leading to the generation of a primary-amine-type weakly-basic anion exchange resin. The increase in the U(VI) capacity would be due to the formation of the straight-chain groups. However, such structural changes were not observed by heating at 323 K in up to 6 M HNO₃ for 20 days. Thus, it was shown that PVPP is a heat-resistant resin.     

Preparation and characterization of an extraction chromatography column for technetium separation based on Aliquat-336 and silica gel support

An extraction chromatographic material based on Aliquat-336 anchored on hydrophobized silica gel support was prepared as an ion exchanger. The prepared material appeared to be suitable for the separation of ⁹⁹Tc from environmental matrices in column application. The properties of the material, sorption characteristic and distribution coefficient of ^99mTcO₄ ^-in various media were studied. The prepared sorbent was conditioned by washing with nitric acid. The solution containing ^99mTcO₄ ^- in 0.1M HNO₃ was passed through the column. Tc was eluted from the column by 8M HNO₃. The flow rate was 0.4 ml/min. The chemical yield of technetium during the separation process was determined using ^99mTc tracer and gamma measurement. The sorption recovery of Tc from the prepared sorbent with 0.1M HNO₃ solution was more than 98%, and the desorption recovery from the column using 8M HNO₃ varied between 92-96%. It was found that the prepared sorbent is suitable for the separation of technetium from environmental matrices and radioactive wastes. This revised version was published online in July 2006 with corrections to the Cover Date.     

Determination of trace amounts of plutonium in low-active liquid wastes from spent nuclear-fuel reprocessing plants by flow injection-based solid-phase extraction/electrochemical detection system

A flow injection-based electrochemical detection system coupled to a solid-phase extraction column was developed for the determination of trace amounts of plutonium in low-active liquid wastes from spent nuclear-fuel reprocessing plants. The oxidation state of plutonium in a sample solution was adjusted to Pu(VI) by the addition of silver(II) oxide. A sample solution was made up in 3 mol L^?1 HNO₃ and loaded onto a column packed with UTEVA^® with 3 mol L^?1 HNO₃ as the carrier. Plutonium(VI) was adsorbed onto the resin, and interfering elements were removed by rinsing the column with 3 mol L^?1 HNO₃. Subsequently, the adsorbed Pu(VI) was eluted with 0.01 mol L^?1 HNO₃, and then introduced directly into the flow-through electrolysis cell with boron-doped diamond electrode. The eluted Pu(VI) was detected by an electrochemical amperometric method at a working potential of 0.1 V (vs. Ag/AgCl). The current produced on reduction of Pu(VI) was continuously monitored and recorded. The plutonium concentration was calculated from the relationship between the peak area and concentration of plutonium. The relative standard deviation of ten analyses was 1.1% for a plutonium solution of 25 μg L^?1 containing 50 ng of Pu. The detection limit calculated from three-times the standard deviation was 0.82 μg L^?1 (1.6 ng of Pu).     

Determination of neptunium in a reprocessing plant by an on-line solid-phase extraction/electrochemical detection system

In this study, a flow-based electrochemical detection system coupled to a solid-phase extraction column was developed for the determination of neptunium in the presence of Pu(IV). Np(V) in the sample solution was completely oxidized to Np(VI) via electrolysis using a column electrode composed of carbon fibers. The column electrode effluent was then loaded onto a TEVA^® column, and subsequently onto a UTEVA^® column using 3 mol L^?1 HNO₃. Pu(IV) was retained on the TEVA column and separated from Np(VI), while Np(VI) was retained on the UTEVA column. Np(VI) was eluted from the UTEVA column with 0.01 mol L^?1 HNO₃ and then introduced directly into a flow-through electrolysis cell. An electrochemical amperometric method with a working potential of +0.1 V (vs. Ag/AgCl) was used to detect Np(VI). The current produced due to the reduction of Np(VI) was continuously monitored and recorded, and the Np concentration was calculated from the peak area. The relative standard deviation of 10 analyses was 2.4 % for an Np solution (0.50 mg L^?1) containing 1.0 μg Np. The detection limit, which was determined to be three times the standard deviation, was 35 μg L^?1 (70 ng Np).     

Recovery of plutonium from phosphate containing aqueous analytical waste solutions using macroporous anion exchange resin

Distribution ratios of Pu(IV) between 7.5M HNO₃+0.75M H₃PO₄+0.3M H₂SO₄ media and a macroporous anion-exchange resin Amberlyst A-26 (MP) increased from 40 to 250 when 1M aluminium nitrate was added to the aqueous medium. When 1M ferric nitrate was used in place of aluminium nitrate the distribution ratio further increased to 850. The 10% Pu(IV) breakthrough capacities with a 5 ml bed resin column, using synthetic feed solutions containing 1M aluminium nitrate, were 1.4 g l^–1, 3.2 g l^–1 at flow rates of 30 ml per hour and 10 ml per hour, respectively. The corresponding 10% Pu(IV) breakthrough capacities in the presence of 1M ferric nitrate were 8.5 g l^–1 and 12.8 g l^–1. More than 97% of plutonium could be recovered from actual analytical phosphate waste solutions.