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.     

PC-controlled radiochemistry system for preparation of no-carrier-added <Superscript>64</Superscript>Cu

A homemade modular PC-controlled radiochemistry system, used to separate the no-carrier-added ⁶⁴Cu from irradiated electroplated solid cyclotron target ⁶⁴Ni layers was developed. Upon irradiation the target layer is dissolved in diluted nitric acid, followed by selective liquid–liquid extraction of the ⁶⁴Cu. The final purification step is achieved making use of an anion exchange column. All the separation procedures can be carried out remote-controlled with a minimized risk of operator errors within less than 2 h, with excellent processing yield (>95%).     

Production and separation of no-carrier-added <Superscript>208,209,210</Superscript>At produced from heavy ion activation on natural thallium target

No-carrier-added (nca) ^208,209,210At was produced for the first time from ⁹Be induced reaction on thallium carbonate target at BARC-TIFR pelletron, Mumbai, India. The target of 4 mg/cm² thickness was prepared by centrifugation technique. Nca At was separated from the thallium target by liquid–liquid extraction using liquid cation exchanger HDEHP dissolved in cyclohexane and liquor ammonia.     

A comparative study on the separation of radioyttrium from Sr- and Rb-targets via ion-exchange and solvent extraction techniques,with special reference to the production of no-carrier-added <Superscript>86</Superscript>Y, <Superscript>87</Superscript>Y and <Superscript>88</Superscript>Y using a cyclotron

The radiochemical separation of ⁸⁸Y from proton irradiated ^natSrCO₃ and alpha-particle irradiated ^natRbCl, of ⁸⁶Y from proton irradiated ⁸⁶SrCO₃, and of ⁸⁷Y from alpha-particle irradiated ^natRbCl were studied at no-carrier-added levels by two techniques, namely, ion-exchange chromatography using Dowex 50W-X8 and Dowex 21K resins, and solvent extraction using HDEHP. Out of all those methods, the ion-exchange chromatography using Dowex 50W-X8 (cation-exchanger) was found to be the best: the separation yield was high, the chemical impurity in the separated radioyttrium (inactive Sr or Rb) was low (0.5 μg) and the final product was obtained in the form of citrate. The optimized separation method using Dowex 50W-X8 was applied in practical production of ⁸⁶Y and ⁸⁸Y via proton irradiations of ⁸⁶SrCO₃ and ^natSrCO₃, respectively, at 16 MeV as well as of ⁸⁷Y and ⁸⁸Y via α-particle irradiation of ^natRbCl at 26 MeV. The tangible experimental yields of ⁸⁶Y and ⁸⁷Y amounted to 150 and 5.7 MBq/μA·h, respectively. The yields of ⁸⁸Y obtained were 0.06 MBq/μA·h and 1 MBq/μA·h for alpha-particle and proton irradiations, respectively. Each yield value corresponds to more than 70% of the respective theoretical value.     

Cyclotron production of <Superscript>169</Superscript>Yb: a potential radiolanthanide for brachytherapy

In the present study, ytterbium-169 was produced via the ¹⁶⁹Tm(p, n)¹⁶⁹Yb nuclear process at the AMIRS (Cyclone-30, IBA, Belgium) cyclotron, irradiating Tm₂O₃ with proton particles of 15 MeV primary energy and 20 μA current for 20 min. Deposition of Tm₂O₃ on Cu substrate was carried out via by the sedimentation method. The 543 mg of thulium(III)oxide with 108 mg of ethyl cellulose and 8 mL of acetone were used to prepare a Tm₂O₃ layer of 11.69 cm². Yields of about 0.643 MBq ¹⁶⁹Yb per μAh were experimentally obtained. ¹⁶⁹Yb was separated in 80 ± 5% radiochemical yield using liquid–liquid extraction. Solvent extraction of no-carrier-added ¹⁶⁹Yb from irradiated thulium(III)oxide target hydrochloric solution was investigated using di-(2-ethylhexyl)phosphoric acid (HDEHP).     

Determination of <Superscript>237</Superscript>Np, <Superscript>93</Superscript>Zr and other long-lived radionuclides in medium and low level radioactive waste samples

The majority of long-lived radionuclides produced in the nuclear fuel cycle can be regarded as “difficult-to-measure” nuclides, hence chemical separation is needed before the nuclear measurement of them. A combined radiochemical procedure that enables the simultaneous determination of some “difficult-to-measure” nuclides in medium and low level radioactive wastes has been developed in our laboratory. Recently, this method has been extended for determination of ²³⁷Np and ⁹³Zr. ²³⁷Np and ⁹³Zr are pre-concentrated by co-precipitation on iron(II) hydroxide and zirconium oxide, separated by extraction chromatography using UTEVA, and measured by inductively coupled plasma mass spectrometry (ICP-MS). As even traces of polyatomic ions and isotopes at m/z 237 or 93 cause considerable interferences during ICP-MS detection, a purification step by extraction chromatography was needed. Analyzing real samples (evaporation concentrates of a nuclear power plant) 66–99% and 31–99% chemical yields were achieved for Np and Zr, respectively.     

A selective separation method for <Superscript>93</Superscript>Zr in radiochemical analysis of low and intermediate level wastes from nuclear power plants

The zirconium isotope ⁹³Zr is a long-lived pure β-particle-emitting radionuclide produced from ²³⁵U fission and from neutron activation of the stable isotope ⁹²Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, ⁹³Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of ⁹³Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. A radiochemical procedure based on liquid–liquid extraction with 1-(2-thenoyl)-3,3,3-trifluoroacetone in xylene, ion exchange with Dowex resin and selective extraction using TRU resin has to be carried out in order to separate zirconium from the matrix and to analyze it by liquid scintillation spectrometry technique (LSC). To set up the radiochemical separation procedure for ⁹³Zr, a tracer solution of ⁹⁵Zr was used in order to follow the behavior of zirconium during the process by γ-ray spectrometry through measurement of the ⁹⁵Zr. Then, the protocol was applied to low level waste (LLW) and intermediate level waste (ILW) from nuclear power plants. The efficiency detection for ⁶³Ni was used to determination of ⁹³Zr activity in the matrices analyzed. The limit of detection of the 0.05 Bq l⁻¹ was obtained for ⁶³Ni standard solutions by using a sample:cocktail ratio of 3:17 mL for OptiPhase HiSafe 3 cocktail.     

Technology of DTPA and immunoglobulins conjugation and their attachment to <Superscript>90</Superscript>Y and <Superscript>177</Superscript>Lu radionuclides

The aim of the study of labeling of ligand–antibody conjugates was to find optimal conditions of preparing of these conjugates and appropriate radioactivity of selected nuclide for applications in nuclear medicine. Conjugation of the γ-immunoglobulin G (human or bovine IgG, polyclonal antibodies) and bifunctional chelating agent, diethylenetriaminepentaacetic acid dianhydride (cDTPAA), was carried out. Various values of the cDTPAA/antibody ratio, the weight concentration of polyclonal or monoclonal antibodies (MEM-97) and buffers were used. Further, the labeling conditions of the DTPA–IgG conjugate by radionuclides ⁹⁰Y and ¹⁷⁷Lu were optimized, and the labeling yield and the conjugation ratio of prepared radionuclide–DTPA–IgG conjugates was determined. Optimal incubation time of the immunoglobulin conjugation was obtained at 30 min from mixing of individual components. The labeling yield of radionuclide–DTPA–antibody conjugate higher than 95% was achieved. Higher values of conjugation ratio of radionuclide–DTPA–antibody conjugate were achieved in 0.1 mol L⁻¹ carbonate buffer, pH 8.5, and the 0.1 mol L⁻¹ carbonate buffer is suitable for studied conjugation systems. This study showed that the labeling yield as well as the conjugation ratio of tested systems depend on the amount of antibody substance, bifunctional chelating agent/antibody molar ratio and pH value of the buffer used.     

Ion-exchange chromatography using citrate buffer,EDTA and α-HIBA for improving the separation of no-carrier-added <Superscript>139</Superscript>Ce from La<Subscript>2</Subscript>O<Subscript>3</Subscript> target irradiated with 14.5 MeV protons

The adsorption behaviour of La/Ce system on Dowex 50W-X8 in different media, namely, nitric acid, acetate buffer and citrate buffer was studied as a function of the concentration of nitric acid and buffer pH. In addition, in cation-exchange column chromatography experiments, three different eluants, namely, citrate buffer of pH 5.5, 0.1 M EDTA and 0.2 M α-HIBA, were employed for separation of Ce(III) from La(III). The optimum conditions for improvement of radiochemical separation of no-carrier-added ¹³⁹Ce from proton irradiated lanthanum were applied using the most suitable chelating agent 0.2 M α-HIBA. The purification of ¹³⁹Ce from macro amount of La(III) was done using two columns in a sequence. The target was prepared by pressing. The production of high radionuclidic and chemical purity ¹³⁹Ce via irradiation of lanthanum oxide target at MGC-20 cyclotron of proton energy 14.5 MeV was described. The experimental yield was found to be 200 kBq/μA h.     

Preparation of <Superscript>26</Superscript>Al, <Superscript>59</Superscript>Ni, <Superscript>44</Superscript>Ti, <Superscript>53</Superscript>Mn and <Superscript>60</Superscript>Fe from a proton irradiated copper beam dump

The station for pions cancer therapy was operated at PSI from 1980 to 1992. After a cooling time of 12 years it’s made of copper beam dump was cut and samples were taken for analytical purposes. The sampling collected about 500 g of high active copper chips that can be used for separation of exotic radionuclides. The analyses by gamma spectrometry, LSC and AMS showed main nuclides present to be ⁶⁰Co, ⁵⁴Mn, ²²Na, ⁶⁵Zn, ²⁶Al, ⁵³Mn, ⁵⁹Ni, ⁶³Ni, ⁵⁵Fe and ⁶⁰Fe and ⁴⁴Ti with a daughter nuclide ⁴⁴Sc. In the frame of ERAWAST project a procedure combining selective precipitation and ion exchange for the separation of the rare radionuclides from the copper beam dump was developed. The proposed separation procedure is easy for remote controlled implementation in a hot cell. The ion exchange separation of Ni, Al, Mg, Ti and Fe was complete and high decontamination factors for copper and cobalt were achieved. Based on the developed procedure a remotely controlled system for separation of exotic radionuclides from the copper chips was set up. The full scale system was installed in a hot cell where high activity levels can be handled. In order to evaluate the reliability and functionality of the system extensive tests have been done. During the test period 13.86 g in total of the proton irradiated copper beam dump were processed for separation of ²⁶Al, ⁵⁹Ni, ⁵³Mn, ⁴⁴Ti and ⁶⁰Fe. The results showed that the system was operational and the radionuclide separation was selective with high chemical yield. The procedure manages as well the generated liquid wastes containing high level of ⁶⁰Co activity.     

Polymer-coated magnetic nanoparticles for rapid bioassay of <Superscript>90</Superscript>Sr in human urine samples

A rapid bioassay for ⁹⁰Sr was developed involving preconcentration of ⁹⁰Sr/⁹⁰Y from human urine samples with a cation exchange polymer (poly–acrylamido–methyl–propanesulfonic acid) coated onto magnetic nanoparticles, followed by selective elution of ⁹⁰Sr (over ⁹⁰Y) with phosphate for determination by liquid scintillation analysis. The minimum detectable activity for this method (4.9 ± 0.5 Bq/L) is lower than the required sensitivity of 19 Bq/L for ⁹⁰Sr in human urine samples, as defined in the requirements for radiation emergency bioassay techniques for the public and first responders based on the dose threshold for possible medical attention recommended by the International Commission on Radiological Protection. The relative bias was 9.2%, the relative precision was 3.2%, and the linear dynamic range covered 12–600 Bq/L. This simple and rapid bioassay method is found to be in compliance with the HPS ANSI N13.30 performance criteria for radiobioassay.     

Study of the cyclotron production of <Superscript>172</Superscript>Lu: an excellent radiotracer

¹⁷²Lu due to its suitable (T _1/2 = 6.7 days) and high detection sensitivity, is used as a radiotracer in different fields. ¹⁷²Lu appears to be suitable as a long-lived rare-earth tracer for compound labelling and biodistribution studies. In the present study, excitation functions via ¹⁷²Yb(p,n)¹⁷²Lu, ^natYb(p,xn)¹⁷²Lu, ¹⁷²Yb(d,2n)¹⁷²Lu and ^natYb(d,xn)¹⁷²Lu reactions were calculated by ALICE/91, ALICE/ASH and TALYS-1.2 codes. Deposition of ^natYb₂O₃ on Cu substrate was carried out via sedimentation method for the production of ¹⁷²Lu. Cementation separation process and liquid–liquid extraction (LLX) of no-carrier-added (nca) radiolutetium from irradiated ytterbium(III)oxide target hydrochloric solution was described using Na(Hg) amalgam, α-hydroxyisobutyric acid (α-HIB) and di-(2-ethylhexyl)phosphoric acid (HDEHP).     

Separation of no-carrier-added <Superscript>90</Superscript>Nb from proton induced natural zirconium target

Amongst various radionuclides of molybdenum, ⁹⁰Mo and ⁹⁹Mo have suitable β energy for clinical uses. In this paper we report separation of ⁹⁹Mo from ⁹⁹Mo-^99mTc equilibrium mixture. The liquid–liquid extraction technique has been employed using trioctylamine (TOA) diluted in cyclohexane as organic phase and HCl as aqueous phase. At 10⁻⁵ M HCl and 0.5 M TOA concentration ^99mTc quantitatively transferred to the organic phase leaving ⁹⁹Mo in the aqueous phase. The developed separation method is efficient and provides very high separation factor.     

Determination of <Superscript>151</Superscript>Sm and <Superscript>147</Superscript>Pm using liquid scintillation tracer methods

The long-lived rare earth isotopes ¹⁵¹Sm (90 years, β _max = 76.3 keV) and ¹⁴⁷Pm (2.62 years, β _max = 224.6 keV) are low-yield fission products that generally require lengthy separation procedures to isolate and count by their beta emissions. We will describe novel liquid scintillation counting techniques using radioactive tracers to determine radiochemical yields from an environmental matrix. The recovery of ¹⁵¹Sm is determined from the alpha decay (2.25 MeV) of ¹⁴⁷Sm in the natural Sm carrier and is in excellent agreement with the gravimetric recovery. The ¹⁴⁷Pm recovery is determined by the use of ¹⁴⁵Pm (17.7 years, EC) tracer, custom-produced at LANL using an isotopically enriched target of ¹⁴⁴Sm. We have determined the ¹⁴⁵Pm recovery both from the 37.4 keV k_α1 X-ray, and the electron-capture emissions by LSC. A comparison of these recovery methods is presented.     

Separation and determination of <Superscript>90</Superscript>Sr in low- and intermediate-level radioactive wastes using extraction chromatography and LSC

A methodology for the determination of ⁹⁰Sr in low- and intermediate-level radioactive wastes from nuclear power plants is presented in this work. It is a part of a methodology developed for the sequential radiochemical separation of radionuclides difficult-to-measure directly by gamma spectrometry in these radioactive wastes. The separation procedure was carried out using precipitation and extraction chromatography with Sr Resin, from Eichrom and the ⁹⁰Sr was measured by liquid scintillation counting (LSC). Optimum conditions for the pretreatment, separation and LSC measurements were determined using simulated samples, which were prepared using standard solutions and carriers. The procedure showed to be rapid and achieved a good chemical yield, in the range 60–90%, and a detection limit of 6.0 × 10⁻⁴ Bq g⁻¹. The method was also tested by participation in a national intercomparison program, with aqueous samples, with good agreement of results.     

Radionuclide purity assessment and calibration of <Superscript>90</Superscript>Y(<Superscript>177</Superscript>Lu) glass particles using gamma-ray spectrometry

Intra-hepatic administration of radioactive glass microspheres is a treatment for patients with primary liver cancer and hepatic metastases. The purpose of this study was radionuclide purity assessment of new glass particles containing two radionuclide, ⁹⁰Y as a therapeutic source and also ¹⁷⁷Lu as a source of diagnostic gamma. For the mixed source, activity measurement using a dose calibrator cannot be used and we need new calibration methods. YAS (Yb) and YAS compositions were sol–gel derived glass particles and production of ⁹⁰Y (¹⁷⁷Lu) and ⁹⁰Y particles was performed using the Tehran Research Reactor. The radionuclide purity was carried out using γ-spectrometry with HPGe detector. A non-destructive spectroscopic assay was employed due to a newly updated low uncertainty positron branching ratio of ⁹⁰Y that emit 511 keV annihilation radiations. In another method, a new calibration of ⁹⁰Y using a non-destructive spectroscopic assay of ⁸⁸Y were investigated. Potential radionuclide impurity include: ⁸⁸Y, ¹⁵²Eu, ⁶⁰Co with activity 100, 50 and 5 Bq per 1 mg of that are not harmful for patients due to delivering radioactive particles about 20–50 mg in ⁹⁰Y(¹⁷⁷Lu) glass microspheres. Among of radionuclide impurity, ¹⁵²Er with a half life of 13.54 years and ⁸⁸Y with a half life of 106.65 days was important in the residual delivery device. For calibration of ⁹⁰Y with monitoring of 511 keV, errors were12.2–21%. In calibration of ⁹⁰Y using gamma spectroscopic assay of ⁸⁸Y, there was an error less than 14%. Spectroscopic assay of ⁸⁸Y can be performed easily and has more repeat for our purpose.     

Determination of <Superscript>3</Superscript>H and <Superscript>14</Superscript>C in the Dry Active Wastes (DAW) generated from the Light Water Reactors (LWR) by a wet oxidation method

³H and ¹⁴C Measurements of the dry active waste (DAW), such as the cotton, paper, and vinyl, generated from a nuclear power plant (NPP) were conducted with wet oxidation using open vessel equipment based on simulation results. The recovery efficiency with the simulated samples was around 93% with a relative standard deviation (RSD) of 1–3%. A liquid scintillation counter (LSC) was used for counting and adjusted to a quenching correction curve. The counting value was evaluated for the minimum detectable activity (MDA), which was found to be about 4 × 10⁻¹ Bq/g for ³H and 2 × 10⁻² for ¹⁴C when approximately 5 g of the samples were measured. The measured DAW samples for the cotton, paper, and vinyl generated from NPP achieved of RSD values of 25, 25, and 60%, respectively, for ³H and 0–50% for ¹⁴C.     

Quinolinephosphomolybdate as ion exchanger: synthesis,characterization, and application in separation of <Superscript>90</Superscript>Y from <Superscript>90</Superscript>Sr

An inorganic ion exchanger, quinolinephosphomolybdate has been synthesized and characterized by elemental analysis, infrared (IR) and X-ray diffraction (XRD) spectroscopy. This compound is highly stable toward thermal, chemical and radiation dose. This has been employed in the separation of carrier-free ⁹⁰Y from its parent ⁹⁰Sr from an equilibrium mixture. The absorbed daughter was recovered by using 0.0284 mol L⁻¹ ascorbic acid solutions at pH 5.0 as eluting agent.     

Separation of <Superscript>113m</Superscript>In from <Superscript>113</Superscript>Sn based on activated carbon used as column matrix

The activated carbon was prepared by using corncobs and characterized by sorpatometer for using as an exchanger material to separate the generated ^113mIn from ¹¹³Sn and ^124,125Sb. To optimize the separation process, the different parameters like acetone percentage, HCl concentration were studied. The exchange capacity of Sn(IV) is 7.6 meq/g onto the activated carbon and the elution efficiency of ^113mIn > 80% by using 10 mL of 0.2 M HCl-80% acetone with flow rate 1 mL/min. The radionuclidic purity and radiochemical purity of the eluted ^113mIn were examined and clarified the presence of ^124,125Sb with relatively high level as radio impurities, so further separation was carried out by using Dowex 1×8 as an anion exchanger below the activated carbon matrix on the same separation column to adsorb the ¹¹³Sn and ^124,125Sb, which escape from the activated carbon matrix.     

Vertical inventories and fluxes of <Superscript>210</Superscript>Pb, <Superscript>228</Superscript>Ra and <Superscript>226</Superscript>Ra at southern South China Sea and Malacca Straits

Inventories and fluxes of ²¹⁰Pb, ²²⁸Ra and ²²⁶Ra were determined in sediment cores collected at nine stations covering of the southern South China Sea and Malacca Straits with the thickness of water column between 42 and 83 m depth. The inventories of ²¹⁰Pb, ²²⁸Ra and ²²⁶Ra were calculated range from 0.15–2.55 Bq cm⁻², 0.05–0.40 Bq cm⁻² and 6.83–83.63 Bq cm⁻², meanwhile the fluxes ranged from 0.005–0.079 Bq cm⁻² yr⁻¹, 0.009–0.048 Bq cm⁻² yr⁻¹ and 0.003–0.037 Bq cm⁻² yr⁻¹, respectively. The results show that the highest inventories and fluxes for ²¹⁰Pb, ²²⁸Ra and ²²⁶Ra were found at station WC 01 and EC 05. Because there are additional sources of ²¹⁰Pb, ²²⁸Ra and ²²⁶Ra, where water transport will brings more dissolved isotopes, influence of the transportation and deposition of suspended particles, fast rate of regeneration and greater production of those radionuclides and others.