A compact solvent extraction based 99Mo/99mTc generator for hospital radiopharmacy.

A compact and portable 99Mo-99 mTc generator based on solvent-extraction, mimic to the conventional 99Mo-99 mTc alumina column generator is much-needed commodity for use in hospital radiopharmacy setup. The present study includes the development of a portable, simple and low cost 99Mo/99 mTc-generator based on MEK solvent extraction technique to obtain a very high concentration of no-carrier added (nca) 99 mTc solution, where low specific activity 99Mo source is obtained through 98Mo(n, γ)99Mo reaction in a research reactor. The unit is intended for operation under the conditions of medical radiological laboratories. Technical trials showed that the mean time of preparation of sodium [99mTc] pertechnetate radiopharmaceutical did not exceed 15 min. The quality and yield of 99 mTc-pertechnetate is upto the mark for formulation of radiopharmaceuticals.

Reductive immobilization of pertechnetate in soil and groundwater using synthetic pyrite nanoparticles.

Radioactive technetium (99Tc) is of intense concern because of its toxicity and high mobility in the environment. Reduction of Tc(VII) to Tc(IV) decreases the mobility and availability of technetium in soil and groundwater. In this study, pyrite nanoparticles (FeS2) were synthesized, characterized and tested for immobilizing/removing 99Tc(VII) in soil and groundwater through batch and column experiments. Influences of particle dosage, dissolved organic matter (DOM), and pH on the reductive immobilization kinetics were examined. At a dosage of 0.28 g/L as Fe, the pyrite nanoparticles were able to rapidly and completely remove 4.88 × 10-7 M of Tc(VII) by converting it to insoluble Tc(IV), with a retarded first-order rate constant of 0.30 h-1. The presence of high concentrations of DOM only moderately inhibited the reduction effectiveness, and acidic pH was more favorable for Tc(VII) reduction. Column experiments showed that embedding a 0.8 cm pyrite layer of the material in a soil bed, simulating a permeable reactive barrier, was able to retard technetium transport 710 times more than a model sandy soil. The results demonstrated that the pyrite particles may serve as a long-lasting reactive material to remediate Tc-contaminated soil, groundwater and solid wastes.

Selective Removal of Technetium from Water Using Graphene Oxide Membranes.

The effective removal of radioactive technetium ((99)Tc) from contaminated water is of enormous importance from an environmental and public health perspective, yet many current methodologies are highly ineffective. In this work, however, we demonstrate that graphene oxide membranes may remove (99)Tc, present in the form of pertechnetate (TcO4(-)), from water with a high degree of selectivity, suggesting they provide a cost-effective and efficient means of achieving (99)Tc decontamination. The results were obtained by quantifying and comparing the free energy changes associated with the entry of the ions into the membrane capillaries (ΔFperm), using molecular dynamics simulations. Initially, three capillary widths were investigated (0.35, 0.68, and 1.02 nm). In each case, the entry of TcO4(-) from aqueous solution into the capillary is associated with a decrease in free energy, unlike the other anions (SO4(2-), I(-), and Cl(-)) investigated. For example, in the model with a capillary width of 0.68 nm, ΔFperm(TcO4(-)) = -6.3 kJ mol(-1), compared to ΔFperm(SO4(2-)) = +22.4 kJ mol(-1). We suggest an optimum capillary width (0.48 nm) and show that a capillary with this width results in a difference between ΔFperm(TcO4(-)) and ΔFperm(SO4(2-)) of 89 kJ mol(-1). The observed preference for TcO4(-) is due to its weakly hydrating nature, reflected in its low experimental hydration free energy.

Removal of TcO4(-) ions from solution: materials and future outlook.

Technetium mainly forms during artificial nuclear fission; it exists primarily as TcO4(-) in nuclear waste, and it is among the most hazardous radiation-derived contaminants because of its long half-life (t1/2 = 2.13 × 10(5) years) and environmental mobility. The high water solubility of TcO4(-) (11.3 mol L(-1) at 20 °C) and its ability to readily migrate within the upper layer of the Earth's crust make it particularly hazardous. Several types of materials, namely resins, molecular complexes, layered double hydroxides, and pure inorganic and metal-organic materials, have been shown to be capable of capturing TcO4(-) (or other oxoanions) from solution. In this review, we give a brief description about the types of materials that have been used to capture TcO4(-) and closely related oxyanions so far and discuss the possibility of using metal-organic frameworks (MOFs) as next-generation ion-exchange materials for the stated application. In particular, with the advent of ultra-stable MOF materials, in conjunction with their chemical tunability, MOFs can be applied to capture these oxyanions under real-life conditions.

A Water-Stable Cationic Metal-Organic Framework as a Dual Adsorbent of Oxoanion Pollutants.

A three-dimensional water-stable cationic metal-organic framework (MOF) pillared by a neutral ligand and with Ni(II)  metal nodes has been synthesized employing a rational design approach. Owing to the ordered arrangement of the uncoordinated tetrahedral sulfate (SO4 (2-) ) ions in the channels, the compound has been employed for aqueous-phase ion-exchange applications. The compound exhibits rapid and colorimetric aqueous-phase capture of environmentally toxic oxoanions (with similar geometries) in a selective manner. This system is the first example of a MOF-based system which absorbs both dichromate (Cr2 O7 (2-) ) and permanganate (MnO4 (-) ) ions, with the latter acting as a model for the radioactive contaminant pertechnetate (TcO4 (-) ).

Separation of [(99m)Tc]pertechnetate and molybdate using polyethylene glycol coated C18 and C30 resins.

Hydrophobic adsorbents such as C18 and C30 were coated with PEG and subsequently used for the separation of Mo/Tc. The most effective resin for adsorbing PEG was the C18-U resin, which demonstrated a coating capacity of 97.6±2.8mg PEG per g of resin. The ability to adsorb pertechnetate was proportional to the amount of PEG coated on the hydrophobic resin. The [(99m)Tc]pertechnetate recovery during the separation of cyclotron produced (99m)Tc from (100)Mo was 91.8±0.3% (n=2). The resultant product met relevant USP monograph specifications.

Radioisotopic Purity of Sodium Pertechnetate 99mTc Produced with a Medium-Energy Cyclotron: Implications for Internal Radiation Dose, Image Quality, and Release Specifications.

UNLABELLED: Cyclotron production of 99mTc is a promising route to supply 99mTc radiopharmaceuticals. Higher 99mTc yields can be obtained with medium-energy cyclotrons in comparison to those dedicated to PET isotope production. To take advantage of this capability, evaluation of the radioisotopic purity of 99mTc produced at medium energy (20-24 MeV) and its impact on image quality and dosimetry was required. METHODS: Thick 100Mo (99.03% and 99.815%) targets were irradiated with incident energies of 20, 22, and 24 MeV for 2 or 6 h. The targets were processed to recover an effective thickness corresponding to approximately 5-MeV energy loss, and the resulting sodium pertechnetate 99mTc was assayed for chemical, radiochemical, and radionuclidic purity. Radioisotopic content in final formulation was quantified using γ-ray spectrometry. The internal radiation dose for 99mTc-pertechnetate was calculated on the basis of experimentally measured values and biokinetic data in humans. Planar and SPECT imaging were performed using thin capillary and water-filled Jaszczak phantoms. RESULTS: Extracted sodium pertechnetate 99mTc met all provisional quality standards. The formulated solution for injection had a pH of 5.0-5.5, contained greater than 98% of radioactivity in the form of pertechnetate ion, and was stable for at least 24 h after formulation. Radioisotopic purity of 99mTc produced with 99.03% enriched 100Mo was greater than 99.0% decay corrected to the end of bombardment (EOB). The radioisotopic purity of 99mTc produced with 99.815% enriched 100Mo was 99.98% or greater (decay corrected to the EOB). The estimated dose increase relative to 99mTc without any radionuclidic impurities was below 10% for sodium pertechnetate 99mTc produced from 99.03% 100Mo if injected up to 6 h after the EOB. For 99.815% 100Mo, the increase in effective dose was less than 2% at 6 h after the EOB and less than 4% at 15 h after the EOB when the target was irradiated at an incident energy of 24 MeV. Image spatial resolution and contrast with cyclotron-produced 99mTc were equivalent to those obtained with 99mTc eluted from a conventional generator. CONCLUSION: Clinical-grade sodium pertechnetate 99mTc was produced with a cyclotron at medium energies. Quality control procedures and release specifications were drafted as part of a clinical trial application that received approval from Health Canada. The results of this work are intended to contribute to establishing a regulatory framework for using cyclotron-produced 99mTc in routine clinical practice.

Cross-linked polyethylene glycol beads to separate 99mTc-pertechnetate from low-specific-activity molybdenum.

UNLABELLED: We report a kit-based approach for the purification of sodium pertechnetate ((99m)TcO4 (-)) from solutions with high MoO4 (2-) content. METHODS: Cross-linked polyethylene glycol resins (ChemMatrix) were used to separate (99m)Tc and molybdenum in 4N NaOH. The resins were loaded at various flow rates and eluted with water to release (99m)Tc. The (99m)Tc solution was passed through a cation exchange resin and an alumina cartridge, followed by saline elution. This process was tested with cyclotron-produced (99m)Tc using an automated system and disposable kits. RESULTS: Optimal results were obtained by loading 500 mg of resin at flow rates of up to 3.1 mL/min, with quantitative extraction of (99m)Tc from the molybdate solution and complete release of (99m)Tc after elution with water. The automated system was highly efficient at isolating Na(99m)TcO4 within minutes, with a recovery rate of 92.7% ± 1.1% (mean ± SD) using cyclotron-produced (99m)Tc. CONCLUSION: ChemMatrix resins were highly effective at separating (99m)TcO4 (-) from molybdate solutions.

Evaluation of silica/ferrocyanide composite as a dual-function material for simultaneous removal of ¹³7Cs⁺ and 99TcO4⁻ from aqueous solutions.

A novel mesoporous silica-coated ferrocyanide (MSCFC) composite was successfully synthesized and evaluated as a dual-function material for simultaneous removal of (137)Cs(+) cations and (99)TcO4(-) anions from aqueous solutions. Sorption behavior of both radionuclides on MSCFC under different experimental conditions has been studied using a batch technique. Results revealed that about 100% of (137)Cs(+) and 97% of (99)TcO4(-) were removed by MSCFC in the pH ranges of 2.2-12.4 and 4.1-9.5, respectively. Sorption kinetic data were analyzed by pseudo-first-order, pseudo-second-order and intraparticle diffusion kinetic models, while Langmuir and Freundlich models were applied for the sorption isotherms. The maximum sorption capacity of MSCFC for radiocesium was determined and compared with other reported sorbents. Applicability of the coated ferrocyanide for simultaneous removal of (137)Cs(+) and (99)TcO4(-) from low-level liquid radioactive waste (LLLW) was also tested, and the data revealed that 99.91% and 98.34% were removed from (137)Cs(+) and (99)TcO4(-), respectively. It is concluded that MSCFC exhibits excellent efficiency for simultaneous removal of the mixed radionuclides with different charge from LLLW.

Implementation of Multi-Curie Production of (99m)Tc by Conventional Medical Cyclotrons.

UNLABELLED: (99m)Tc is currently produced by an aging fleet of nuclear reactors, which require enriched uranium and generate nuclear waste. We report the development of a comprehensive solution to produce (99m)Tc in sufficient quantities to supply a large urban area using a single medical cyclotron. METHODS: A new target system was designed for (99m)Tc production. Target plates made of tantalum were coated with a layer of (100)Mo by electrophoretic deposition followed by high-temperature sintering. The targets were irradiated with 18-MeV protons for up to 6 h, using a medical cyclotron. The targets were automatically retrieved and dissolved in 30% H2O2. (99m)Tc was purified by solid-phase extraction or biphasic exchange chromatography. RESULTS: Between 1.04 and 1.5 g of (100)Mo were deposited on the tantalum plates. After high-temperature sintering, the (100)Mo formed a hard, adherent layer that bonded well with the backing surface. The targets were irradiated for 1-6.9 h at 20-240 µA of proton beam current, producing up to 348 GBq (9.4 Ci) of (99m)Tc. The resulting pertechnetate passed all standard quality control procedures and could be used to reconstitute typical anionic, cationic, and neutral technetium radiopharmaceutical kits. CONCLUSION: The direct production of (99m)Tc via proton bombardment of (100)Mo can be practically achieved in high yields using conventional medical cyclotrons. With some modifications of existing cyclotron infrastructure, this approach can be used to implement a decentralized medical isotope production model. This method eliminates the need for enriched uranium and the radioactive waste associated with the processing of uranium targets.

An automated module for the separation and purification of cyclotron-produced 99mTcO4-.

INTRODUCTION: The shortage of reactor-produced molybdenum-99 ((99)Mo, t(½)=66 h) has renewed interest in alternative production methods of its daughter isotope, technetium-99m ((99m)Tc, t(½)=6.02 h). While adsorption chromatography serves as a mechanism for selective elution of sodium pertechnetate from technetium generators, this method of purification is not sufficient for many alternative production methods. Several ion-separation/solid phase extraction chromatography methods are known, yet none have been demonstrated on cyclotron-produced [(99m)Tc]TcO(4)(-). Herein we describe the design, manufacture and optimization of a remotely operated module for the purification of sodium pertechnetate from a bulk solution of molybdate. METHODS: The automated purification module was designed to separate [(99m)Tc]TcO(4)(-) using either Dowex 1x8 or an Aqueous Biphasic Extraction Chromatography (ABEC) resin. (100)Mo composite targets were irradiated with 18.5 MeV protons for 10 µA·h using an ASCI TR19 cyclotron. Once purified, the radiopharmaceutical quality of (99m)TcO(4)(-) isolated from each process (Dowex and/or ABEC) was established by assaying for molybdate breakthrough, alumina levels and, in the case of the Dowex approach, residual organics. RESULTS: The separation processes are efficient (75% for Dowex, 90% for ABEC) and complete in less than 30 min. Overall, up to 2.1 GBq of (99m)Tc was produced using the (100)Mo(p,2n)(99m)Tc transformation, processed using the separation module and subjected to a detailed chemical and radionuclidic analysis. Due to its expense and limited availability, (100)MoO(4)(2-) was recovered in >90% yield using a precipitation/filtration/lyophilization approach. CONCLUSIONS: Na[(99m)Tc]TcO(4) was produced using a medical cyclotron, recovered using an automated purification module and found to exceed all established quality control parameters.

Anion separation and preconcentration with cyclen and cyclen-resorcinarene derivatives.

N-undecyl cyclen and a resorcinarene bowl bearing four cyclen arms have been applied as anion exchangers in ion chromatography by strong adsorption to a reversed-phase column. The column loaded with the resorcinarene bowl cyclen tetramer exhibited significantly better performance in anion separation than that with N-undecyl cyclen monomer in isocratic elution mode. Both columns were tested for polarizable anion preconcentration or removal. By changing the eluent type from sodium bicarbonate to sodium hydroxide, the degree of protonation of the cyclen molecules could be modified, and the column capacity for anion retention adjusted thereby. Capacity gradient elution was successfully applied to removing sample matrix ions in the preconcentration of perchlorate and perrhenate ions as example analytes.

Molecular recognition of pertechnetate and perrhenate.

This tutorial review covers the chemistry of artificial receptors designed to recognize the pertechnetate and perrhenate anions. It focuses on the molecular features that give rise to effective and selective complexation. Specific receptor classes are discussed in the context of available solution-phase and solid-state data with differences between pertechnetate and perrhenate binding behavior being highlighted where appropriate.

Sodium sulfate as an eluant for concentrated solution of pertechnetate.

Fission molybdenum-99 based technetium-99m chromatography generators are the most widely employed generator systems in the field of diagnostic nuclear medicine. In the first week these generators provide high specific volume activity Na99mTcO4, but in the second week the radioactive concentration of the eluate (99mTc) becomes 17% of that at the beginning of the first week. Low specific volume activity limits the clinical procedures (e.g. 99mTc-DTPA for lung ventilation scanning, 99mTc-sestamibi for myocardial perfusion studies, 99mTc-ECD for brain imaging and 99mTc-red cells for testicles) that can be performed at the beginning of the second week. To overcome such limitations sodium sulfate was used as an eluant, which provides>80% of 99mTc radioactivity in 2 ml solution, while 0.9% NaCl provides only 40-45% 99mTc in first 2 ml solution.

Effect of Zr:Mo ratio on 99mTc generator performance based on zirconium molybdate gels.

Zirconium molybdate gels have shown to be viable alternatives for preparation of 99mTc generators using 99Mo produced by neutron activation. The aim of this work was to investigate the effect of the Zr:Mo molar ratio on the gel chemical structure and correlate it with the elution efficiency. A series of gels were prepared at Zr:Mo molar ratios from 0.5:1 to 2.3:1 and characterized by TGA, IR, XRD and UV. It was found that the variation of Zr:Mo ratio produces different polymolybdate arrangements on the octahedral units around to the zirconia which is mainly influenced by the water content. When the matrix molybdenum concentration was increased a lesser amount of water was found and the elution efficiencies were increased. However high elution efficiencies produce higher 99Mo breakthrough values. The gel formulation appears thus to be a compromise between the elution efficiency and the molybdenum breakthrough. The chemical-physical properties of these gels are presented and discussed.

A rapid and stable ITLC procedure for the determination of the radiochemical purity of 99Tcm-tetrofosmin.

99Tcm-tetrofosmin (Myoview, Amersham Healthcare) is widely used as a radiopharmaceutical for myocardial perfusion imaging. Control of its radiochemical purity after reconstitution is usually performed by means of ITLC-SG paper chromatography in a mobile phase of methylene chloride/acetone (65/35), as recommended by the manufacturer. The present study describes the application of tetrahydrofuran in phosphate buffer for the development of 99Tcm-tetrofosmin on ITLC-SG strips. The use of this mobile phase significantly improves the separation between labelled tetrofosmin and unbound pertechnetate. The time for development is about 1 min and the solvent is stable for at least 1 year. In addition, the volume spotted on the strip does not affect the migration of 99Tcm-tetrofosmin. The labelling efficiency of 99Tcm-tetrofosmin can successfully be monitored by means of this method as a daily routine procedure.

Retardation capacity of organophilic bentonite for anionic fission products.

Sorption and diffusivity of iodide and pertechnetate (I- and TcO4-) on MX-80 bentonite with different hexadecylpyridinium (HDPy+) loadings were studied using equilibrium solutions of different ionic strengths. In HDPy(+)-modified bentonite, iodide and pertechnetate ions exhibited increasing sorption (characterized by the distribution ratio, Rd), while Cs+ and Sr2+ showed decreasing sorption with increasing organophilicity. In case of medium-loading levels, the simultaneous sorption of anions (I- and TcO4-) and cations (Cs+ and Sr2+) was observed. Sorption of ions was influenced by the composition of the electrolytes employed. It decreased gradually with increasing ionic strength of the electrolyte solutions. The experiments revealed the general tendency that the diffusivity (Da [cm2.s-1]) for iodide and pertechnetate decreases with increasing organophilicity and increases with increasing ionic strength of the equilibrium solutions, confirming the results of the sorption experiments. Additionally, some mineralogical and chemical investigations, like IR spectral analysis of the organo-bentonite samples and exchange behavior of HDPy+, were performed. On the basis of these analyses, it was concluded that the alkylammonium ions are sorbed as (1) HDPy+ cations, (2) HDPyCl molecules and (3) micelles with decreasing binding intensities in this order.

Plasma clearance and biodistribution of oxidatively modified 99mTc-beta-VLDL in rabbits.

The biodistribution and removal from plasma (measured as fractional clearance rate, FCR, per hour) of native and oxidatively modified 99mtechnetium-labeled beta-very low density lipoprotein (99mTc-beta-VLDL) were investigated in hypercholesterolemic (HC) and control (C) three-month old New Zealand rabbits. The intracellular accumulation of beta-VLDL labeled with 99mTc was studied in vitro in THP-1 cells and monocyte-derived macrophages isolated from rabbits. After intravenous injection into C rabbits, copper-oxidized beta-VLDL (99mTc-ox-beta-VLDL) was cleared from the circulation faster (0.362 +/- 0.070/h) than native beta-VLDL (99mTc-nat-beta-VLDL, 0.241 +/- 0.070/h). In contrast, the FCR of 99mTc-ox-beta-VLDL in HC rabbits was lower (0.100 +/- 0.048/h) than that of 99mTc-nat-beta-VLDL (0.163 +/- 0.043/h). The hepatic uptake of radiolabeled lipoproteins was lower in HC rabbits (0.114 +/- 0.071% injected dose/g tissue for 99mTc-nat-beta-VLDL and 0.116 +/- 0.057% injected dose/g tissue for 99mTc-ox-beta-VLDL) than in C rabbits (0.301 +/- 0.113% injected dose/g tissue for 99mTc-nat-beta-VLDL and 0.305 +/- 0.149% injected dose/g tissue for 99mTc-ox-beta-VLDL). The uptake of 99mTc-nat-beta-VLDL and 99mTc-ox-beta-VLDL by atherosclerotic aorta lesions isolated from HC rabbits (99mTc-nat-beta-VLDL: 0.033 +/- 0.012% injected dose/g tissue and 99mTc-ox-beta-VLDL: 0.039 +/- 0.017% injected dose/g tissue) was higher in comparison to that of non-atherosclerotic aortas from C rabbits (99mTc-nat-beta-VLDL: 0.023 +/- 0.010% injected dose/g tissue and 99mTc-ox-beta-VLDL: 0.019 +/- 0.010% injected dose/g tissue). However, 99mTc-nat-beta-VLDL and 99mTc-ox-beta-VLDL were taken up by atherosclerotic lesions at similar rates. In vitro studies showed that both monocyte-derived macrophages isolated from rabbits and THP-1 macrophages significantly internalized more 99mTc-ox-beta-VLDL than 99mTc-nat-beta-VLDL. These results indicate that in cholesterol-fed rabbits 99mTc-ox-beta-VLDL is slowly cleared from plasma and accumulates in atherosclerotic lesions. However, although the extent of in vitro uptake of 99mTc-ox-beta-VLDL by macrophages was high, the in vivo accumulation of this radiolabeled lipoprotein by atherosclerotic lesions did not differ from that of 99mTc-nat-beta-VLDL.

Plasma clearance and biodistribution of oxidatively modified 99mTc-Beta-VLDL in rabbits

The biodistribution and removal from plasma (measured as fractional clerance rate, FCR, per hour) of native and oxidatively modified (99m)technetium-labeled Beta-very low density lipoprotein ((99m)Tc-Beta-VLDL)) were investigated in hypercholesterolemic (HC) and control (C) three-month old New Zealand rabbits. The intracellular accumulation of Beta-VLDL labeled with (99m)Tc was studied in vitro in THP-1 cells and monocyte-derived macrophages isolated from rabbits. After intravenous injection into C rabbits, copper-oxidized Beta-VLDL ((99m)Tc-ox-Beta-VLDL)) was cleared from the circulation faster (0.362 + 0.070/h) than native Beta-VLDL ((99m)Tc-nat-Beta-VLDL, 0.241 + 0.070/h)). In contrast, the FCR of (99m)Tc-ox-Beta-VLDL in HC rabbits was lower (0.100 + 0.048/h) than that of (99m)Tc-nat-Beta-VLDL (0.163 + 0.043/h). The hepatic uptake of radiolabeled lipoproteins was lower in HC rabbits (0.114 + 0.071 percent injected dose/g tissue for (99m)Tc-nat-Beta-VLDL and 0.116 + 0.057 percent injected dose/g tissue for (99m) Tc-ox-Beta-VLDL) than in C rabbits (0.301 + 0.113 percent injected dose/g tissue for (99m)Tc-nat-Beta-VLDL and 0.305 + 0.149 percent injected dose/g tissue for ((99m)Tc-ox-Beta-VLDL). The uptake of (99m)Tc-nat-Beta-VLDL and (99m)Tc-ox-Beta-VLDL by atherosclerotic aorta lesions isolated from HC rabbits ((99m)Tc-nat-Beta-VLDL:0.033 + 0.012 percent injected dose/g tissue and (99m)Tc-ox-Beta-VLDL: 0.039 + 0.017 percent injected dose/g tissue) was higher in comparison to that of non-atherosclerotic aortas from C rabbits (99m)Tc-nat-Beta-VLDL: 0.023 + 0.010 percent injected dose/g tissue and (99m)Tc-ox-Beta VLDL: 0.019 + 0.010 percent injected dose/g tissue). However, (99m) Tc-nat-Beta-VLDL and (99m)Tc-ox-Beta-VLDL were taken up by atherosclerotic lesions at similar rates. In vitro studies showed that both monocyte-derived macrophages isolated from rabbits and THP-1 macrophages significantly internalized more (99m)Tc-ox-Beta-VLDL than (99m)Tc-nat-Beta-VLDL. These results indicate that in cholesterol-fed rabbits (99m)Tc-ox-Beta-VLDL is slowly cleared from plasma and accumulates in atherosclerotic lesions. However, although the extent of in vitro uptake of (99m)Tc-ox-Beta-VLDL by macrophages was high, the in vivo accumulation of this radiolabeled lipoprotein by atherosclerotic lesions did not differ from that of (99m)Tc-nat-Beta-VLDL.

Electronic interaction chromatography on porous graphitic carbon. Separation of [99mTc]pertechnetate and perrhenate anions.

The oxo anions of technetium-99m and rhenium, pertechnetate (TcO4-) and perrhenate (ReO4-), have been separated by high performance liquid chromatography on porous graphitic carbon with aqueous trifluoroacetic acid or salt solutions as eluent. Chromatographic retention was a result of electronic interaction between the lone pair electrons of the anions and the delocalised pi-electron clouds of the porous graphitic carbon. Retention and resolution can be controlled by the concentration and/or species of eluents which can compete with the solutes for electronic interaction being used for elution.