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.

Fractionated elution of 99Mo/99mTc generators using calibrated vials.

Fractionated elution consists in collecting the fractions of an eluate with the highest radioactive concentration. It may be useful to meet the requirements of a subset of clinical radiopharmacy procedures. This study aims to describe and evaluate straightforward procedures allowing to readily perform fractionated elution on dry and wet columns Mo/Tc generators by using calibrated vials. The main objectives of this study consisted in determining the relationship between eluate volume and elution yield using different vials calibration and assessing repeatability of the procedure. Elution vials were calibrated to obtain different eluate volumes by addition of air for wet column generator (WCG) and subtraction of saline for dry column generator (DCG) (n≥5 for each calibration). The relationship between the eluate volume and the elution yield was determined by a regression model for both DCG and WCG. Then repeatability evaluation was performed using 3-ml vial calibration. Relationships between the eluate volume (V) and the elution yield (Y) for DCG and WCG were Y=57.551 ln(V)+10.526 and Y=50.256 ln(V)+17.597, respectively. For repeatability assessment (n=30 for DCG and n=31 for WCG), the median volume and the interquartile range for DCG and WCG were 2.98 ml (2.92-3.01) and 3.28 ml (2.71-3.40), respectively, and median (interquartile range) eluate yields were 84.73% (81.30-86.33) and 81.78% (78.91-85.20), respectively. The volume was significantly higher for WCG than DCG (P=0.036) and also significantly more variable (P<0.001). The elution yield was significantly lower for WCG than DCG (P=0.025), but no difference in variability between the two generators was found (P=0.874). Easy-to-handle fractionated elution methods are compatible with both DCG and WCG. Fractionation using calibrated vials exhibits a better reproducibility with DCG than WCG generators and represents the only proposed method so far to master fractionated elution with DCG.

In-house cyclotron production of high-purity Tc-99m and Tc-99m radiopharmaceuticals.

In the last years, the technology for producing the important medical radionuclide technetium-99m by cyclotrons has become sufficiently mature to justify its introduction as an alternative source of the starting precursor [99mTc][TcO4]- ubiquitously employed for the production of 99mTc-radiopharmaceuticals in hospitals. These technologies make use almost exclusively of the nuclear reaction 100Mo(p,2n)99mTc that allows direct production of Tc-99m. In this study, it is conjectured that this alternative production route will not replace the current supply chain based on the distribution of 99Mo/99mTc generators, but could become a convenient emergency source of Tc-99m only for in-house hospitals equipped with a conventional, low-energy, medical cyclotron. On this ground, an outline of the essential steps that should be implemented for setting up a hospital radiopharmacy aimed at the occasional production of Tc-99m by a small cyclotron is discussed. These include (1) target production, (2) irradiation conditions, (3) separation/purification procedures, (4) terminal sterilization, (5) quality control, and (6) Mo-100 recovery. To address these issues, a comprehensive technology for cyclotron-production of Tc-99m, developed at the Legnaro National Laboratories of the Italian National Institute of Nuclear Physics (LNL-INFN), will be used as a reference example.

Cyclotron production of 99mTc: Comparison of known separation technologies for isolation of 99mTc from molybdenum targets.

Intensive efforts were undertaken during the last few decades for the separation of cyclotron-produced 99mTc from 99Mo and new papers have been published on this topic since the last review [1]. In the future the cyclotron-based methods can replace reactor-based technology in producing this medical radioisotope and the nuclear reaction 100Mo(p,2n)99mTc appears to be the most worthwhile approach. New ways of producing of 99mTc require efficient separation methods. Several strategies for separation of 99mTc from 99Mo have been already developed. The advantages, disadvantages and technical challenges toward application potential of investigated methods to separate 99mTc from irradiated 100Mo target are discussed. These methods include column chromatography, solvent extraction, chemical precipitation and thermochromatography.

Accelerator-based production of the (99m)Tc-(186)Re diagnostic-therapeutic pair using metal disulfide targets (MoS2, WS2, OsS2).

Novel, natural abundance metal disulfide targets were irradiated for 1h with a 10µA proton beam in a small, medical cyclotron. Osmium disulfide was synthesized by simple distillation and precipitation methods while MoS2 and WS2 were commercially available. The targets dissolved under mild conditions and were analyzed by γ-spectroscopy. Production rates and potential applications are discussed, including target recovery and recycling schemes for OsS2 and WS2.

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.

Redox Interactions of Tc(VII), U(VI), and Np(V) with Microbially Reduced Biotite and Chlorite.

Technetium, uranium, and neptunium are contaminants that cause concern at nuclear facilities due to their long half-life, environmental mobility, and radiotoxicity. Here we investigate the impact of microbial reduction of Fe(III) in biotite and chlorite and the role that this has in enhancing mineral reactivity toward soluble TcO4(-), UO2(2+), and NpO2(+). When reacted with unaltered biotite and chlorite, significant sorption of U(VI) occurred in low carbonate (0.2 mM) buffer, while U(VI), Tc(VII), and Np(V) showed low reactivity in high carbonate (30 mM) buffer. On reaction with the microbially reduced minerals, all radionuclides were removed from solution with U(VI) reactivity influenced by carbonate. Analysis by X-ray absorption spectroscopy (XAS) confirmed reductive precipitation to poorly soluble U(IV) in low carbonate conditions and both Tc(VII) and Np(V) in high carbonate buffer were also fully reduced to poorly soluble Tc(IV) and Np(IV) phases. U(VI) reduction was inhibited under high carbonate conditions. Furthermore, EXAFS analysis suggested that in the reaction products, Tc(IV) was associated with Fe, Np(IV) formed nanoparticulate NpO2, and U(IV) formed nanoparticulate UO2 in chlorite and was associated with silica in biotite. Overall, microbial reduction of the Fe(III) associated with biotite and chlorite primed the minerals for reductive scavenging of radionuclides: this has clear implications for the fate of radionuclides in the environment.

Studies on the separation of 99mTc from large excess of molybdenum.

BACKGROUND: Due to aging and unexpected prolonged shutdown of nuclear reactors producing 99Mo for 99Mo/ 99mTc generators it was necessary to explore the alternative methods of technetium-99m production. The first choice were the accelerators. Three years ago IAEA (International Atomic Energy Agency) initiated the Coordinated Research Project "Accelerator-based Alternatives to Non-HEU production of Mo-99 /Tc-99m" aimed at direct production of 99mTc in proton accelerators using the ¹°°Mo(p,2n)99mTc reaction. POLATOM is participating in this enterprise together with the Heavy Ion Laboratory of Warsaw University and the Institute of Nuclear Chemistry and Technology. MATERIAL AND METHODS: 99Mo/99mTc solutions and pure 99mTc used for generators production or milked from ready to use generators were used in experiments. Commercial chromatographic and laboratory-prepared columns were used for separation. The peristaltic pumps were used for solutions delivery onto the columns. Radioactivity of eluted 99Mo and 99mTc was measured using high resolution gamma spectrometry or ionisation chamber in case of high radioactivity. For separation, three different chromatographic methods were used, one based on ion exchange and two on extraction. RESULTS: Synthetic mixtures simulating the real solutions were used. 99mTc is quantitatively bound in the Dowex-1 × 8 column whereas molybdenum is only slightly retained and totally rinsed with 2M NaOH. 99mTc is eluted with TBAB. The elution yield has been reproducible and amounted to 78%. The AnaLig Tc-02 resin column was used for 99mTc retention. Residual Mo was removed by rinsing with 2M NaOH and 99mTc eluted using small volume of water. The recovery was equal to about 85%. Using C-18 column coated with PEG over 80% of 99mTc was recovered in about 50 mL of water. The reduction of volume was necessary. CONCLUSIONS: The recovery of 99mTc was the highest using AnaLig Tc-02 resin. Time of 99mTc separation is the shortest for AnaLig Tc-02 resin and it is not higher than 100 minutes and it can further be shortened.

Development of a freeze-dried kit formulation for the preparation of 99mTc-NTP 15-5, a radiotracer for scintigraphic imaging of proteoglycans.

The cartilage-targeting strategy is based on the strong affinity of quaternary ammonium (QA) functions for cartilage proteoglycans. We use a bifunctional agent containing QA moiety and a polyazamacrocycle structure able to complex technetium-99m. (99m)Tc-NTP 15-5 was selected for its high stability and its high affinity for proteoglycans in vivo. Labeling conditions of NTP 15-5 were optimized, and a lyophilized kit was developed for radiolabeling of (99m)Tc-NTP 15-5 (radiochemical yields 94.6±1.8%). (99m)Tc-NTP 15-5 was stable and resulted in favorable biological evaluations.

Automation of 99Tc extraction by LOV prior ICP-MS detection: application to environmental samples.

A new, fast, automated and inexpensive sample pre-treatment method for (99)Tc determination by inductively coupled plasma-mass spectrometry (ICP-MS) detection is presented. The miniaturized approach is based on a lab-on-valve (LOV) system, allowing automatic separation and preconcentration of (99)Tc. Selectivity is provided by the solid phase extraction system used (TEVA resin) which retains selectively pertechnetate ion in diluted nitric acid solution. The proposed system has some advantages such as minimization of sample handling, reduction of reagents volume, improvement of intermediate precision and sample throughput, offering a significant decrease of both time and cost per analysis in comparison to other flow techniques and batch methods. The proposed LOV system has been successfully applied to different samples of environmental interest (water and soil) with satisfactory recoveries, between 94% and 98%. The detection limit (LOD) of the developed method is 0.005 ng. The high durability of the resin and its low amount (32 mg), its good intermediate precision (RSD 3.8%) and repeatability (RSD 2%) and its high extraction frequency (up to 5 h(-1)) makes this method an inexpensive, high precision and fast tool for monitoring (99)Tc in environmental samples.

Measured bremsstrahlung photonuclear production of 99Mo ((99m)Tc) with 34 MeV to 1.7 GeV electrons.

(99)Mo photonuclear yield was measured using high-energy electrons from Laser Plasma Accelerators and natural molybdenum. Spectroscopically resolved electron beams allow comparisons to Monte Carlo calculations using known (100)Mo(γ,n)(99)Mo cross sections. Yields are consistent with published low-energy data, and higher energy data are well predicted from the calculations. The measured yield is (15±2)×10(-5) atoms/electron (0.92±0.11 GBq/µA) for 25 mm targets at 33.7 MeV, rising to (1391±20)×10(-5) atoms/electron (87±2 GBq/µA) for 54 mm/ 1.7 GeV, with peak power-normalized yield at 150 MeV.

Diversification of 99Mo/99mTc separation: non­fission reactor production of 99Mo as a strategy for enhancing 99mTc availability.

This paper discusses the benefits of obtaining (99m)Tc from non-fission reactor-produced low-specific-activity (99)Mo. This scenario is based on establishing a diversified chain of facilities for the distribution of (99m)Tc separated from reactor-produced (99)Mo by (n,γ) activation of natural or enriched Mo. Such facilities have expected lower investments than required for the proposed chain of cyclotrons for the production of (99m)Tc. Facilities can receive and process reactor-irradiated Mo targets then used for extraction of (99m)Tc over a period of 2 wk, with 3 extractions on the same day. Estimates suggest that a center receiving 1.85 TBq (50 Ci) of (99)Mo once every 4 d can provide 1.48-3.33 TBq (40-90 Ci) of (99m)Tc daily. This model can use research reactors operating in the United States to supply current (99)Mo needs by applying natural (nat)Mo targets. (99)Mo production capacity can be enhanced by using (98)Mo-enriched targets. The proposed model reduces the loss of (99)Mo by decay and avoids proliferation as well as waste management issues associated with fission-produced (99)Mo.

The excitation functions of (100)Mo(p,x)(99)Mo and (100)Mo(p,2n)(99m)Tc.

Proton-induced nuclear reactions for generation of (99)Mo and (99m)Tc radionuclides were investigated using the stacked-foil activation technique on 99.05% enriched (100)Mo targets at energies up to Ep=21MeV. Excitation functions of the reactions (100)Mo(p,x)(99)Mo and (100)Mo(p,2n)(99m)Tc have been measured.

Incorporation and retention of 99-Tc(IV) in magnetite under high pH conditions.

Technetium incorporation into magnetite and its behavior during subsequent oxidation has been investigated at high pH to determine the technetium retention mechanism(s) on formation and oxidative perturbation of magnetite in systems relevant to radioactive waste disposal. Ferrihydrite was exposed to Tc(VII)(aq) containing cement leachates (pH 10.5-13.1), and crystallization of magnetite was induced via addition of Fe(II)aq. A combination of X-ray diffraction (XRD), chemical extraction, and X-ray absorption spectroscopy (XAS) techniques provided direct evidence that Tc(VII) was reduced and incorporated into the magnetite structure. Subsequent air oxidation of the magnetite particles for up to 152 days resulted in only limited remobilization of the incorporated Tc(IV). Analysis of both X-ray absorption near edge structure (XANES) and extended X-ray absorption fine structure (EXAFS) data indicated that the Tc(IV) was predominantly incorporated into the magnetite octahedral site in all systems studied. On reoxidation in air, the incorporated Tc(IV) was recalcitrant to oxidative dissolution with less than 40% remobilization to solution despite significant oxidation of the magnetite to maghemite/goethite: All solid associated Tc remained as Tc(IV). The results of this study provide the first direct evidence for significant Tc(IV) incorporation into the magnetite structure and confirm that magnetite incorporated Tc(IV) is recalcitrant to oxidative dissolution. Immobilization of Tc(VII) by reduction and incorporation into magnetite at high pH and with significant stability upon reoxidation has clear and important implications for limiting technetium migration under conditions where magnetite is formed including in geological disposal of radioactive wastes.

Evaluation of alternative rapid thin layer chromatography systems for quality control of technetium-99m radiopharmaceuticals.

Whatman 3MM™ and Tec-Control™ systems were evaluated as ITLC-SG alternatives for 99mTc-radiopharmaceuticals. They compare well in accuracy and reproducibility, and are faster and more convenient than ITLC-SG. Tec-Control™ radiochemical purity values for 99mTc-sestamibi were more conservative than ITLC-SG. Full solvent migration was not reproduced for 99mTc-tetrofosmin in Tec-Control™, and for this Whatman 3MM™ is preferred. Developing times were 10-15 min, 7-9 min and ~1min for ITLC-SG, Whatman 3MM™ and Tec-Control™, respectively. Overall, Tec-Control™ strips are preferred due to speed and ease of use.

Prediction of 94mTc production for positron emission tomography studies using the Monte Carlo code MCNPX-2.6.

In this study, a Monte Carlo code was used to simulate a proton beam flux to calculate the (94m)Tc production yield from the (94)Mo(p,n)(94m)Tc reaction. An experimental yield of 3.465 GBq/µAh was measured for 48 min of irradiation at 1µA. An estimated value of 3.048 GBq/µAh was calculated for the yield produced based on the MCNPX proton flux in the same situation. These results demonstrate the usefulness and precision of MCNPX as a tool to design targets for the production of PET radionuclides. The yield of isotopic impurities from the (94)Mo(p,xn)(94g, 93m, 93g)Tc reactions was also calculated by the MCNPX code.

99Mo/(99m)Tc separation: an assessment of technology options.

Several strategies for the effective separation of (99m)Tc from (99)Mo have been developed and validated. Due to the success of column chromatographic separation using acidic alumina coupled with high specific activity fission (99)Mo (F (99)Mo) for production of (99)Mo/(99m)Tc generators, however, most technologies until recently have generated little interest. The reduced availability of F (99)Mo and consequently the shortage of (99)Mo/(99m)Tc column generators in the recent past have resurrected interest in the production of (99)Mo as well as (99m)Tc by alternate routes. Most of these alternative production processes require separation techniques capable of providing clinical grade (99m)Tc from low specific activity (99)Mo or irradiated Mo targets. For this reason there has been renewed interest in alternate separation routes. This paper reviews the reported separation technologies which include column chromatography, solvent extraction, sublimation and gel systems that have been traditionally used for the fabrication of (99)Mo/(99m)Tc generator systems. The comparative advantage, disadvantage, and technical challenges toward adapting the emerging requirements are discussed. New developments such as solid-phase column extraction, electrochemical separation, extraction chromatography, supported liquid membrane (SLM) and thermochromatographic techniques are also being evaluated for their potential application in the changed scenario of providing (99m)Tc from alternate routes. Based on the analysis provided in this review, it appears that some proven separation technologies can be quickly resurrected for the separation of clinical grade (99m)Tc from macroscopic levels of reactor or cyclotron irradiated molybdenum targets. Furthermore, emerging technologies can be developed further to respond to the expected changing modes of (99m)Tc production.

Sorption coefficients and molecular mechanisms of Pu, U, Np, Am and Tc to Fe (hydr)oxides: a review.

Pu, U, Np, Am and Tc are among the major risk drivers at nuclear waste management facilities throughout the world. Furthermore, uranium mining and milling operations have generated an enormous legacy of radioactively contaminated soils and groundwater. The sorption process of radionulcides onto ubiquitous Fe (hydr)oxides (FHOs; hematite, magnetite, goethite and ferrihydrite) is one of the most vital geochemical processes controlling the transport and fate of radionuclides and nuclear wastes in the subsurface zones. Meanwhile, understanding molecular-level chemical speciation of radionuclides onto FHOs is crucial to model their behavior in subsurface environments, and to develop new technologies for nuclear waste treatment and long-term remediation strategies for contaminated soils and groundwater. This review article aims (1) to provide risk or performance assessment modelers with macroscopic distribution coefficient (K(d)) data of Pu, U, Np, Am and Tc onto FHOs under different conditions (pH, radionuclide concentration, solution ion strength, sorbent loading, partial pressure of CO(2) (P CO(2)), equilibrium time) pertinent to environmental and engineered systems, and (2) to provide a microscopic or molecular-level understanding of the chemical speciation and sorption processes of these radionuclides to FHOs.

A computerized compact module for separation of (99m)Tc-radionuclide from molybdenum.

An automated closed cycle module for the separation and recovery of various isotopes, radioactive or non-radioactive, using solvent extraction and column chromatography techniques, and in particular, for separation and recovery of (99m)Tc from low-medium specific activity (99)Mo, is described. The module may also be applicable for separation of (99m)Tc produced in a cyclotron. The module is safe and reliable to avoid human interference and hazards posed by handling of radioactive and hazardous chemicals. The entire system of automation includes a user-friendly PC based graphical user interface (GUI) that actually supervises the process via an embedded system based electronic controller.

An electrochemical procedure to concentrate 99mTc availed from a zirconium [99Mo] molybdate gel generator.

Successful use of electrochemical method to concentrate (99m)Tc obtained from a zirconium molybdate (Zr(99)Mo) gel generator is reported. The process consists of electrodeposition of the (99m)Tc in an oxalate bath on a small platinum electrode followed by recovery of (99m)Tc in a small volume of normal saline. The overall recovery of (99m)Tc was >90% with >99.99% radionuclidic purity and >99% radiochemical purity. The compatibility of the product in the preparation of (99m)Tc labeled formulations such as (99m)Tc-DMSA and (99m)Tc-EC was found to be satisfactory in terms of high labeling yields (>98%).