<Superscript>47</Superscript>Sc production development by cyclotron irradiation of <Superscript>48</Superscript>Ca

The therapeutic radionuclide ⁴⁷Sc was produced through the ⁴⁸Ca(p,2n) channel on a proton beam accelerator. The obtained results show that the optimum proton energies are in the range of 24–17 MeV, giving the possibility to produce ⁴⁷Sc radionuclide containing 7.4% of ⁴⁸Sc. After activation, the powdery CaCO₃ target material was dissolved in HCl and scandium isotopes were isolated from the targets. The performed separation experiments indicate that, due to the simplicity of the operations and the chemical purity of the obtained ⁴⁷Sc the best separation process is when scandium radioisotopes are separated on the 0.2 µm filter.     

Production of carrier free <Superscript>188</Superscript>Re radioisotope generator based on aluminum tungstate matrix

Improved radionuclide generator include a substantially insoluble salt of a radioactive parent which may be directly packed in column for subsequent elution of the daughter radionuclide. An improved ¹⁸⁸Re generator was prepared by reacting a radioactive tungsten (¹⁸⁸W) as parent radionuclide incorporated with aluminum chloride to obtain an insoluble radioactive aluminum tungstate matrix. The investigated matrix was characterized on the basis of the chemical composition, IR, thermal analysis and mechanical stabilities. The factors affecting the elution performance were studied such as influence of pH, molar ratio and drying temperature. From the obtained data, the molar ratio W:Al was 1.5:1 at pH = 4, the matrix dried at 105 °C for 2 h. Chromatographic and multichannel analysis has been currently used to investigate the radiochemical and radionuclidic purity respectively on eluted ¹⁸⁸Re. An elution yield more than 80%, with radiochemical purity <98% and radionuclidic purity <99% with a ¹⁸⁸W break through >10⁻⁴% of the column. The Al⁺³ and W contents value were about 2 and 3 μg/mL eluate. The obtained data approved the stability of the prepared generator and its suitability for medical application.     

Production and chemical separation of <Superscript>48</Superscript>V

Summary Vanadium-48 was produced with an activity of 222 MBq (6 mCi) by the ^natTi(d,xn)⁴⁸V nuclear reactions in the U-120 cyclotron. The energy of the irradiating beam was 13 MeV, its intensity 5 μA, and the metallic Ti target dimension 11’11 mm² (0.1 mm thickness). For target cooling, circulated water in target backside was used. After 3 cooling days, only ⁴⁸V and some ⁴⁶Sc (T_1/2 = 83.8 d), produced by the side nuclear reaction ⁴⁸Ti(d,α)⁴⁶Sc were found in the target. For production of the radiotracer of ⁴⁸V and for the preparation of source for standardization of ⁴⁸V by 4πβ-γ coincidence, the Ti target was dissolved either in HF or in H₂SO₄. For both dissolving methods an ion-exchange separation procedure was developed.     

Preparation and measurement of the half-life of <Superscript>101</Superscript>Tc

¹⁰¹Tc is a very important nuclide as fuel burn up monitor, and its half-life value has been measured many times, however, they were so different from each other. In this work, ¹⁰¹Tc liquid samples were prepared by irradiating analytical pure (NH₄)₆Mo₇O₂₄·4H₂O solution in Miniature Neutron Source Reactor (MNSR) of China. A rapid procedure which takes only 5 min was developed to separate ¹⁰¹Tc samples. The final samples were analyzed by γ-ray spectrum using high purity Germanium (HPGe) multi-channel analysis system. The results showed that there were no peaks of other nuclides except ¹⁰¹Tc. The half-life of ¹⁰¹Tc was accurately measured with HPGe γ-detector following 306.8 keV γ-ray for about 140 min, and three methods R-value method, iterative method and translation method were adopted to process the data. Finally, a more precise and accurate value 14.02 ± 0.01 min was given and compared with former measured data.     

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.     

Determination of activity ratios of <Superscript>238,239+240,241</Superscript>Pu, <Superscript>241</Superscript>Am, <Superscript>134,137</Superscript>Cs,and <Superscript>90</Superscript>Sr in Bulgarian soils

The origins of different artificial radionuclides found in soils from Northern and Southern Bulgaria was determined by measurements of their actual concentrations and respective ratios. On the basis of the measured mobility and concentrations of the investigated radionuclides in soils, it was estimated that after the Chernobyl accident the mean depositions of fresh ¹³⁷Cs were 3.0 ± 2.5 kBq/m² for Northern Bulgaria and 15 ± 7 kBq/m² for Southern Bulgaria. As a result of global fallout following atmospheric nuclear weapon tests in the 1950s, mean depositions (corrected to 1965) were calculated for Northern and Southern Bulgaria as follows: for ⁹⁰Sr—1.0 ± 0.5 and 2.3 ± 1.3 kBq/m², ²³⁸Pu—1.3 ± 0.8 and 2.8 ± 1.6 Bq/m², ^239+240Pu—15 ± 14 and 47 ± 38 Bq/m², and ²⁴¹Pu—520 ± 200 and 760 ± 260 Bq/m².     

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.     

Production of monoenergetic MeV-range neutrons by <Superscript>3</Superscript>H(p,n)<Superscript>3</Superscript>He reaction

Monochromatic MeV-energy neutron source for secondary reaction was developed utilizing tritium embedded titanium (Ti-³H) thin film via ³H(p,n)³He reaction. We have measured the neutron energies and the energy spread by resonance reactions of ¹²C(n,tot) and ²⁸Si(n,tot). The available energy was within the range from 0.6 to 2.6 MeV. Energy spread was 1.6% at energy of 2.077 MeV. The flux in the beam direction was determined to be 3.76·10⁷ n/s/sr by irradiating ¹⁹⁷Au by about 2 MeV neutrons. This source was shown to be useful for measurements of nuclear data by measuring the total cross sections of neutrons on Fe and comparing these data to the data of ENDF-6.     

Systematic radiochemical separation for the determination of <Superscript>99</Superscript>Tc, <Superscript>90</Superscript>Sr, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe and <Superscript>59,63</Superscript>Ni in low and intermediate radioactive waste samples

A simple and rapid separation procedure was systemized for the determination of ⁹⁹Tc, ⁹⁰Sr, ⁹⁴Nb, ⁵⁵Fe and ^59,63Ni in low and intermediate level radioactive wastes. The integrated procedure involves precipitation, anion exchange and extraction chromatography for the separation and purification of individual radionuclide from sample matrix elements and from other radionuclides. After separating Re (as a surrogate of ⁹⁹Tc) on an anion change resin column, Sr, Nb, Fe and Ni were sequentially separated as follows; Sr was separated as Sr (Ca-oxalate) co-precipitates from Nb, Fe and Ni followed by purification using Sr-Spec extraction chromatographic resin. Nb was separated from Fe and Ni by anion exchange chromatography. Fe was separated from Ni by anion exchange chromatography. Ni was separated as Ni-dimethylglyoxime precipitates after the removal of ^134,137Cs and ^110mAg by Cs-phosphotungstate and AgCl precipitation, respectively. Finally, the radionuclide sources were prepared by precipitation for their radioactivity measurements. The reliability of the procedure was evaluated by measuring the recovery of chemical carriers added to a synthetic radioactive waste solution.     

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).     

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.     

Average cross-sections for (n, α) and (n,p) reactions on titanium in a fission-type reactor spectrum

The average cross-section in a fission-type reactor spectrum was determined experimentally for the reactions:⁴⁶Ti(n,p)⁴⁶Sc,⁴⁷Ti(n,p)⁴⁷Sc,⁴⁸Ti(n,p)⁴⁸Ti(n,α)⁴⁵Ca and⁵⁰Ti(n,α)⁴⁷Ca. In order to obtain the (n,p) cross-sections, reactor irradiation of titanium was followed by measurement of the induced scandium activities with a Ge(Li) detector of calibrated detection efficiency. For this no chemical separations had to be carried out. For the (n,α) reactions, however, the induced calcium activities were separeted and purified by oxalate precipitation, after the bulk of the radioactivity had been removed by precipitation of titanium hydroxide. The⁴⁷Ca disintegration rate was determined in the same way as for the scandium isotopes, whereas for⁴⁵Ca liquid scintillation counting was carried out. The shape of the reactor spectrum was investigated by irradiating reference threshold detectors with different effective threshold energies. To correct for (n,γ) interferences, irradiations were carried out with and without cadmium shielding. On the basis of \(\bar \sigma _F = 0.64\) mb for the reaction²⁷Al(n,α)²⁴Na, the average cross-sections were as follows:⁴⁶Ti(n,p)⁴⁶Sc:10.5±0.4 mb;⁴⁷Ti(n,p)⁴⁷Sc: 16.3±0.6 mb;⁴⁸Ti(n,p)⁴⁸Sc:0.272±0.005 mb;⁴⁸Ti(n,α)⁴⁵Ca: 34μb;⁵⁰Ti(n,α)⁴⁷Ca: 8.1±0.3 μb.     

Neutron activation analysis of iodine-129 and iodine-127 in environmental samples

The possibilities of reactor induced (n, p) reactions as a tool for neutron activation analysis of titanium in geological samples are discussed. The interference of calcium and scandium is experimentally evaluated. Results for Ti, Ca and Sc in GSP-1 and PCC-1 standard rocks are presented. On the basis of the experimental values, it is concluded that the⁴⁷Ti(n, p)⁴⁷Sc reaction is the most favourable for titanium determination.     

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.     

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.     

The analysis of radiolysis impurities in <Superscript>18</Superscript>F-FDG and methods of repurification

To investigate the radio impurity in the radiolysis of ¹⁸F-FDG at high radiodose and radioconcentrated solutions and develop methods of repurification. The radiolysis of ¹⁸F-FDG was analyzed by TLC. The radio-impurity was confirmed by biodistribution and small animal PET/CT studies. ¹⁸F-FDG was unstable at high radioconcentrition over 37 GBq/mL or under basic condition. TLC, biodistribution and PET/CT all indicated that the main autoradiolysis byproduct was free fluoride ion. The radiolyzed ¹⁸F-FDG was repurified by solid-phase extraction (SPE) column. The repurified ¹⁸F-FDG had a radiochemical purity (RCP) of over 99% and significantly lower bone uptake than that was before repurification (P = 0.0003). There was a positive correlation between the recovery yield and the purity of ¹⁸F-FDG (R ² = 0.66).     

Transport and sorption of <Superscript>85</Superscript>Sr and <Superscript>125</Superscript>I in crushed crystalline rocks under dynamic flow conditions

Transport and sorption of water-soluble ⁸⁵Sr²⁺ and ¹²⁵I⁻ in the columns with beds of crushed crystalline rocks from synthetic groundwater has been studied under dynamic flow conditions. Samples of crystalline rocks: diorite-I, diorite-II, gabbro, granite and tonalite, having the grain size between 0.25 and 0.80 mm, were used. Plastic syringes of 8.8 cm length and 2.1 cm in diameter were applied as columns. The synthetic groundwater was pumped downward through the columns with a seepage velocity of about 0.2 cm/min and the given radioactive nuclide was added into the water stream individually in a form of a short pulse. In case of ⁸⁵Sr, desorption from diorite-I was also studied using an artificial acid rainfall and then, the longitudinal distribution of the residual ⁸⁵Sr activity along the bed was measured. Retardation, distribution and hydrodynamic dispersion coefficients were determined by the evaluation of respective breakthrough curves. A corrected integral form of a simple advection–dispersion equation was derived and used for fitting the experimental data. The K _d-parameters resulting from dynamic experiments were also compared with the results of static sorption experiments.     

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.     

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.     

Determination of <Superscript>40</Superscript>K, <Superscript>226</Superscript>Ra and <Superscript>232</Superscript>Th concentration and radiation dose assessment in titanium enamel frits use in polishing household utensils in Ghana

As part of monitoring the exposure of the Ghanaian public to natural radioactivity, radioactivity concentrations in titanium enamel frits use for coating, spraying and decorating steel bowls were investigated. Samples collected from Ghana Utensil Manufacturing Company in Ghana were analyzed using γ-ray spectrometry with a high purity germanium detector. The samples were found to contain an average absorbed dose rate of 509.38 nGy h⁻¹, while an average annual effective dose was calculated to be 2.50 mSv.