Zirconium silicotungstate matrix as a prospective sorbent material for the preparation of <Superscript>113</Superscript>Sn/<Superscript>113m</Superscript>In generator

The zirconium silicotungstate (ZrSiW) was studied as an effective sorbent material to be used in the ¹¹³Sn/^113mIn generator. The results elucidated that the distribution coefficient of ¹¹³Sn (3700 mL/g) is greater than ^113mIn (275 mL/g) from 0.1 M HCl acid solution to the ZrSiW material. The maximum sorption capacity of Sn (IV) was found to be 33 mg per gram ZrSiW (~?0.3 mmol/g). The elution yield of ^113mIn was found to be >?78?±?6.4% with an acceptable purity of radionuclidic and radiochemical (≥?99.99 and 96.8%, respectively). The rigorous separation of ^113mIn from the ¹²⁵Sb was carried out due to its long half-life (2.758 years) and beta emission that causes tissue damage. Zr, W and Si levels are below the permitted limit in the ^113mIn eluate.     

Potential use of tungstocerate matrices on the separation of <Superscript>113m</Superscript>In from <Superscript>110m</Superscript>Ag relevant to the separation of the cyclotron-produced <Superscript>111</Superscript>In from its silver target

A procedure for separation of no-carrier-added ^113mIn(III) radioisotope from a bulk of ^110mAg has been developed. The sorption behavior of ^113mIn(III) and ^110mAg(I) ions in HNO₃ acid solutions on different tungstocerate matrices showed high affinity of ^110mAg(I) ions towards tungstocerate(IV) gel matrices compared with ^113mIn(III) ions. No-carrier-added ^113mIn radionuclide was separated from ^110mAg on 12-tungstocerate(IV) column matrix. 11 mL 0.3M HNO₃ acid solution was enough for eluting the ^113mIn from the column bed. ^110mAg was recovered from the column by eluting the column bed with 12 mL 2M HNO₃ acid solution.     

<Superscript>137</Superscript>Cs/<Superscript>137m</Superscript>Ba radioisotope generator based on 6-tungstocerate(IV) column matrix

Summary Barium-137m radioisotope generator of the chromatographic column elution mode based on loading 1.5 g 6-tungstocerate(IV) gel matrix with ~54 kBq of fission-produced ¹³⁷Cs is described. The elution performance of the generated ^137mBa radionuclide was investigated as a function of chemical composition of the eluent, flow rate, elution frequency, and age of the generator system. At comparable conditions, ^137mBa eluates with 0.9% NaCl-0.1M HCl eluent had higher elution yields and radionuclidic purity than with 0.1M NH₄Cl-0.1M HCl eluent. The generator has been repeatedly eluted for 311 days by passing 4810 ml of the saline eluent (10 ml × 481 elution operations) at a flow rate of 3.0 ml/min. Barium-137m eluates of high and reproducible elution yields, chemical and radionuclidic purities of (≥ 99.99%) were obtained.     

Separation of no-carrier-added <Superscript>149</Superscript>Gd from <Superscript>12</Superscript>C activated natural praseodymium matrix

¹⁴⁹Gd was produced from the ¹²C induced reaction on natural praseodymium target. No-carrier-added (nca) ¹⁴⁹Gd was separated from the bulk target matrix by liquid–liquid extraction (LLX) using cation exchanger di-(2-ethylhexyl)phosphoric acid (HDEHP) dissolved in cyclohexane. High separation factor of 2,450 was achieved at the optimal experimental condition when 1% HDEHP and 0.1 M HCl were used as organic and aqueous phases respectively. The result was also compared with the previous reports.     

Production of no-carrier-added <Superscript>117m</Superscript>Sn from proton irradiated antimony

A method for production of no-carrier-added ^117mSn (NCA ^117mSn) has been developed. It includes proton irradiation of thick antimony targets and chemical recovery of ^117mSn by extraction of Sb with dibutyl ether and chromatographic purification on silica gel column. The method provides production of curie amounts of ^117mSn with specific activity about 1000 Ci/g and high radionuclidic purity.     

<Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generators based on aluminum molybdate gel matrix prepared by nano method

A procedure for preparation of ⁹⁹Mo/^99mTc radioisotope generator based on low specific activity neutron activated ⁹⁹Mo was developed. Aluminum molybdate(VI)-⁹⁹Mo of high Mo(VI) content (~?364 mg/g Al⁹⁹Mo) was prepared by mixing low specific activity molybdate(VI)-⁹⁹Mo and aluminum mixture solution with isoamyl alcohol. Al⁹⁹Mo gel matrix was precipitated when the pH of the mixture solution was raised to ~?5 by addition of NaOH to the mixture. Radiometric measurements indicate the strong fixation of Molybdate(VI)-⁹⁹Mo species in the form of the sparingly insoluble Al⁹⁹Mo gel matrix. The prepared AlMo gel matrix was physiochemically characterized. Al⁹⁹Mo gel matrix was used as a base material for preparation of ⁹⁹Mo/^99mTc generator. The ^99mTc eluted from ⁹⁹Mo/^99mTc radioisotope generator was found to have relatively high elution yield (84?±?2.3%), radionuclidic (≥?99.99%), radiochemical (98.1?±?0.9%) and chemical purity.     

Development of new <Superscript>116</Superscript>Cd/Pt target for cyclotron produced <Superscript>117m</Superscript>Sn as a medical radiometal

The radioisotope ^117mSn has recently attracted considerable attention because of its application in theranostics and its imaging using 159-keV γ-photons. In this study, we developed a target system that yielded ^117mSn via a ¹¹⁶Cd(α, 3n)^117mSn nuclear reaction. A separable Pt substrate was utilized to prepare the enriched ¹¹⁶Cd target to be irradiated, and the enriched ¹¹⁶Cd in a cadmium acetate solution was electroplated onto the Pt substrate. The substrate was thermally analyzed via ANSYS simulations, and the plating thickness was optimized through calculations with TALYS code. The production of ^117mSn was confirmed through the emission measurement of inherent gamma rays.     

Determination of radionuclidic impurities in <Superscript>99m</Superscript>Tc eluate from <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generator for quality control

Technetium-99m is the principal radioisotope used in medical diagnostics; radionuclidic impurity is the major concern of its quality. This work presents a analytical method for sequential determination of all radionuclidic impurities listed in pharmacopoeia including gamma emitters, alpha emitters, ⁸⁹Sr and ⁹⁰Sr. Radioactive decay for removal of ^99mTc, ion exchange and extraction chromatography for removal of ⁹⁹Mo and ⁹⁹Tc are effective for separation of interferences. Gamma spectrometry, LSC with alpha/beta discrimination, and Cherenkov counting using LSC are sensitive methods for measurement of the impurity radionuclides. The detection limits of this method are well meet the requirement of the quality control according to the limitation of the pharmacopoeia.     

<Superscript>103</Superscript>Ru/<Superscript>103m</Superscript>Rh generator

^103mRh is a very promising radionuclide for Auger electron therapy due to its very low photon/electron ratio. The goal of the present work was the elaboration a method for production of large quantities of ^103mRh for generator system. It was found that the combination of solvent extraction with evaporation of ¹⁰³RuO₄ followed by decomposition of H₅IO₆ makes it possible to produce ^103mRh of high radionuclidic and chemical purity.     

Modeling β-γ coincidence spectra of <Superscript>131m</Superscript>Xe, <Superscript>133</Superscript>Xe, <Superscript>133m</Superscript>Xe,and <Superscript>135</Superscript>Xe

In support of the Comprehensive Nuclear-Test-Ban Treaty (CTBT), improvements have been made to the model of the Automated Radioxenon Sampler/Analyzer (ARSA) β-γ coincidence detector for radioxenon monitoring. MCNPX is used to simulate the detector response for all the electrons and photons emitted from ^131mXe, ¹³³Xe, ^133mXe, ¹³⁵Xe, and ¹³⁷Cs signals. A MatLab code was written to incorporate the MCNPX results in the calculation of β-γ coincidence spectra. These will aid in the development of the Spectral Deconvolution Analysis Tool (SDAT)¹ and to calibrate β-γ coincidence systems. The models developed for this work include improvements over previous models in their ability to address Compton scattering in the β-cell, and the β-distribution offset in the 31 keV γ-ray region for ¹³³Xe.     

Preparation of <Superscript>95m</Superscript>Tc radiotracer

Several procedures for preparation of the ^95mTc radiotracer following irradiation of a thin Mo target with deuterons were tested. The procedures consisting of alkaline-oxidative fusion of the irradiated target in a mixture of Na₂O₂ + NaOH and subsequent liquid–liquid extraction with 2-butanone, and acid decomposition of the target in a mixture of H₂SO₄ + HNO₃ followed by extraction chromatography with PAN-Aliquat 336 composite material appeared suitable for the given purpose.     

A novel electrochemical <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generator

A novel electrochemical process to avail clinical grade ^99mTc from (n,γ)⁹⁹Mo has been demonstrated. The electrochemical parameters were optimized to maximize the ^99mTc yield with minimal ⁹⁹Mo contamination. ⁹⁹Mo/^99mTc generators containing up to 29.6 GBq (800 mCi) ⁹⁹Mo were developed and their performance were extensively evaluated for 10 days without changing the operating conditions. Very high radioactive concentration of ^99mTcO₄ ⁻ of acceptable quality, commensurate with hospital radiopharmacy requirements could be availed from the system with >90% yield. The compatibility of the product for the formulation of ^99mTc labeled radiopharmaceuticals such as ^99mTc-DMSA and ^99mTc-EC was found to be satisfactory in terms of high labeling yields. The proposed route represents an important step for enhancing the scope of accessing clinical grade ^99mTc from low specific activity (n, γ)⁹⁹Mo.     

SDAT: analysis of <Superscript>131m</Superscript>Xe with <Superscript>133</Superscript>Xe interference

The Spectral Deconvolution Analysis Tool (SDAT) software was developed at The University of Texas at Austin. SDAT utilizes a standard spectrum technique for the analysis of β–γ coincidence spectra. Testing was performed on the software to compare the standard spectrum analysis technique with a region of interest (ROI) analysis technique. Experimentally produced standard spectra and sample data were produced at the Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor. The results of the testing showed that the standard spectrum technique had lower errors than the ROI analysis technique for samples with low counting statistics. In contrast, the ROI analysis technique outperformed the standard spectrum technique in high counting statistics samples. It was also shown that the standard spectrum technique benefitted from a compression of the number of channels within the spectra.     

Preparation of <Superscript>99m</Superscript>Tc-metronidazole as a model for tumor imaging

Metronidazole (MTNZ) is an antiprotozoa drug, could be labeled with the ^99mTc. MTZL could be used as an ideal vehicle to deliver radioactive decay energy of ^99mTc to the sites of tumor, thus facilitate tumor imaging. The process of labeling was done using tin chloride as reducing agent. The optimum conditions required to label 25 μg MTZL were 100 μg stannous chloride, 30 min reaction time, room temperature at pH 7–9 using 0.5 M phosphate buffer. The radiochemical purity of the labeled compound, at the above conditions, was determined using paper chromatography. The yield was about 93%. About 2.5 × l0⁶ of Ehrlich Ascites Carcinoma (EAC) was injected intrapritoneally (i.p) to produce ascites and intramuscularly (i.m) in the right thigh to produce solid tumor in female mice. Biodistribution studies were carried out by injecting solution of ^99mTc-MTZL in normal and tumor bearing mice. The uptake in ascites was over 5% of the injected dose per gram tissue body weight, at 4 h post injection and above 4% in solid tumor. These data revealed localization of the tracer in the tumor tissues with high percentage sufficient to use ^99 mTc MTZL as promising tool for diagnosis of tumor.     

Evaluation of fresh and old eluate of <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generators used for labeling of different pharmaceutical kits

Summary Sixty ⁹⁹Mo/^99mTc wet column generators, loaded with two different ⁹⁹Mo activities, were analyzed in order to assess the quality of their eluates. Each elution was used for labeling of different radiopharmaceuticals, in order to evaluate whether “risky” elutions, namely those performed just after generator delivery and at 72 hours or more from the last elution, could be conveniently employed when fresh available radioactivity is not enough for the planned labeling or when shipping problems arise, or delay in delivery of a new generator occurs. Radiochemical quality control of all radiopharmaceuticals labeled with these elutions was performed. The elutions differed mainly in ⁹⁹Tc ground state (^99gTc) and amounts of oxidizing impurities. Radiolabeling procedures, however, were not affected, suggesting that these “risky” elutions might be appropriately used, in “emergency” conditions, for labeling radiopharmaceuticals although their radiochemical purity control is recommended prior to patient administration.     

Effect of humic acid on the sorption of technetium on hematite colloids using <Superscript>95m</Superscript>Tc and <Superscript>96</Superscript>Tc as tracers

Sorption of technetium on hematite colloids, at varying pH (3–10), has been studied in absence and presence of humic acid using ^95mTc-⁹⁶Tc radiotracers. Technetium was found to be weakly sorbed on hematite at lower pH (<5) values, while no sorption was observed at higher pH values. Humic acid was found to have no effect on the sorption of technetium on hematite under aerobic conditions, while at lower pH values small reduction was observed which was attributed to the reduced zeta potential of the hematite colloids owing to the strong sorption of humic acid.     

Separation and purification of <Superscript>249</Superscript>Cf and <Superscript>245</Superscript>Cm

There is a need to provide radioactivity standards of the higher actinides in support of both decommissioning and remediation activities as well as routine environmental analysis. In the case ²⁴⁹Cf, this will provide a useful calibration nuclide for both α-and γ-spectrometry as well as improving knowledge of the decay scheme for this nuclide. There is anecdotal evidence to suggest that the chemical yield of americium and curium may differ in radiochemical analysis. Thus, a chemical yield tracer of ²⁴⁵Cm may help to resolve this issue and will be suitable for both, suitable for use as a chemical yield tracer for both α-particle spectrometry and mass spectrometry. An aged source of ²⁴⁹Cf was used as the source material for the separation of these two nuclides by cation-exchange, using 2-hydroxy-2-methyl-propanoic acid at controlled pH as an eluant, ²⁴⁹Cf being eluted before the ²⁴⁵Cm daughter. The purity of both nuclides was measured by γ-ray spectrometry.     

Selective ion-exchange sorption of K<Superscript>+</Superscript> from aqueous solutions containing Na<Superscript>+</Superscript>-K<Superscript>+</Superscript> mixture on Sn(IV) hydrophosphate

Ion-exchange sorption of K⁺ and Na⁺ from their mixture on X-ray-amorphous and crystalline Sn(IV) hydrophosphates was studied. These sorbents exhibit a high selectivity for K⁺. Amorphous Sn(IV) hydrophosphate can be used for efficient purification of aqueous sodium salt solutions to remove potassium impurity.Translated from Zhurnal Prikladnoi Khimii, Vol. 77, No. 8, 2004, pp. 1275–1278.Original Russian Text Copyright © 2004 by Smirnov, Dimova, Redchenko.     

Combined determination of <Superscript>99</Superscript>Tc and <Superscript>108m</Superscript>Ag in L/ILW liquid wastes

The low- and intermediate-activity level liquid wastes produced by the Paks Nuclear Power Plant (NPP) contain routinely measureable gamma-emitting (e.g., ⁵⁴Mn, ⁶⁰Co, ^110mAg, and ¹³⁷Cs) as well as many so-called “difficult-to-measure” radionuclides. Despite of their low specific activity compared to the total, the reliable determination of these radionuclides is an important issue of nuclear waste management. The increasing amount of waste samples to be qualified yearly by our laboratory put a pressure on revising the existing procedure of ⁹⁹Tc separation applied. We have managed to halve the initial amount of the sample required to achieve the same level of detection of technetium. Furthermore, one of the new purifying steps introduced have proved to be able to separate ^108mAg (and ^110mAg) better than 99% keeping the ⁹⁹Tc content of the product almost intact. Means of separation of ⁹⁹Tc from ¹⁰⁶Ru and ^124+125Sb have also been successfully investigated. As intended, this new procedure has a major impact on the chemical reagent as well as the electricity requirement of the separation making it more cost-effective.