Half-life measurement of <Superscript>126</Superscript>Sn isolated from Hanford nuclear defense waste

Summary Long-term risk assessment of residual and disposed nuclear fuel reprocessing waste requires good knowledge of component isotopes with long half-lives. For example, the accuracy of the accepted ¹²⁶Sn half-life of approximately 100,000 years is insufficient for desired risk assessments. From modeling and sampling, ¹²⁶Sn is known to exist in Hanford nuclear waste. Excess portions of waste characterization samples were used to isolate ¹²⁶Sn for measurement of its half-life. Isolation was performed with ion-exchange resins. The resulting ¹²⁶Sn was gamma-assayed with a hyperpure germanium spectrometer for decay photon identification and activity values. An inductively coupled plasma/mass spectrometer was used to measure the atom quantity of the isolated ¹²⁶Sn. The separation chemistry, observed gamma energies, and calculated half-life are presented. The half-life of ¹²⁶Sn estimated in this work is (2.33±0.10) ^. 10⁵ years.     

Elemental analysis of biological tissues of Dmdmdx/J and C57BL/6J mice strains investigated by neutron activation analysis

Radioactivity measurement of short-lived nuclides is the basis of decay data measurement, which requires rapid separation and purification of the interested nuclides from complicated fission products. A rapid separation system based on SISAK and extraction chromatography technique was established to isolate ⁹⁵Y, which half-life is 10.3 min. With the best conditions studied in this paper, ⁹⁵Y was separated successfully from complicated fission products under the mini-reactor in the China Institute of Atomic Energy. Decontamination factors to other nuclides except ⁹⁴Y are higher than 2 × 10³.     

<Superscript>234</Superscript>U/<Superscript>235</Superscript>U activity ratios as a probe for the <Superscript>238</Superscript>U/<Superscript>235</Superscript>U half-life ratio

High-resolution alpha-particle spectrometry was performed on three uranium materials enriched in ²³⁵U. Besides the ²³⁵U peaks, separate peaks belonging to impurity traces of ²³⁴U could be quantified. Relying on the isotopic composition of the uranium, as determined by mass spectrometry, the ratio of the half-lives of ²³⁸U and ²³⁵U was determined via the activity ratio of ²³⁴U and ²³⁵U in the materials. As an intermediate link, the ²³⁴U/²³⁸U half-life ratio was taken from published mass spectrometric analyses of ‘secular equilibrium’ uranium material. The resulting half-life ratio T _1/2(²³⁸U)/T _1/2(²³⁵U) = 6.351±0.031 is in agreement with the commonly adopted half-life values determined by Jaffey et al.     

Sequential separation of <Superscript>99</Superscript>Tc, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe, <Superscript>90</Superscript>Sr and <Superscript>59/63</Superscript>Ni from radioactive wastes

A sequential separation procedure has been developed for the determination of ⁹⁹Tc, ⁹⁴Nb, ⁵⁵Fe, ⁹⁰Sr and ^59/63Ni in various radioactive wastes generated from nuclear power plants. Ion exchange and extraction chromatography were adopted for individual separation of the radionuclides. Precipitation was supplementarily utilized for both purification of the individual radionuclides and preparation of the radionuclide sources for use in a radioactivity measurement. The chromatographic separation behavior of the radionuclides both from the sample matrix metals and from one another was investigated using stable metals, Re (as a surrogate of ⁹⁹Tc), Nb, Fe, Sr and Ni. The validity of the procedure for reliability and applicability was evaluated by measuring the recovery of the metal carriers added to synthetic radioactive waste solutions. The recoveries by the chromatographic separation were in the range of 84.8 to 102.2% with 2s of less than 8.6%, the recoveries by the precipitation being in the range of 84.3 to 97.3% with 2s of less than 10.9%.     

Atmospheric depositional fluxes of cosmogenic <Superscript>32</Superscript>P, <Superscript>33</Superscript>P and <Superscript>7</Superscript>Be in the Sevastopol region

⁹⁰Y was separated from ⁹⁰Sr using an extraction chromatographic resin consisting of 4, 4′(5′)-bis-t-butylcyclohexano-18-crown-6 (DtBuCH₁₈C₆), 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonyl) amide (C₂mimNTf₂), and a polymer (Amberlite XAD-7). Ionic liquid was introduced into the column to improve the separation efficiency. The column showed an excellent performance for the separation of Y from Sr. After the separation, the ratio of ⁹⁰Sr/⁹⁰Y was <2.0 × 10^?5; the column was recycled for >18 times. This study provides preliminary results on columns to produce ⁹⁰Y with a high purity in radiopharmaceuticals.     

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.     

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.     

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.     

Determination of <Superscript>93</Superscript>Zr in nuclear power plant wastes

A radioanalytical method (based on separation using UTEVA columns and ICP-MS measurement) has been used for determination of ⁹³Zr in 37 nuclear power plant samples. As ⁹³Nb might affect the detection of ⁹³Zr, Monte Carlo activation model was used to calculate the expected ⁹³Zr/^natZr mass ratio, which was compared to measured ones. It was found, that a decontamination factor of 10³ is sufficient to get a measurement uncertainty of less than 10%.     

Extraction chromatography for the separation of <Superscript>239</Superscript>Np from <Superscript>243</Superscript>Am

Summary An extraction chromatography method was developed for the separation of 239Np from ²⁴³Am in nitric acid solution. A sorbent based on aliphatic quaternary amine Aliquat-336 and hydrophobized silica gel was prepared. ²³⁹Np reduced to the oxidation state(IV) with ferrous sulfamate in 2M or 6M HNO₃ sorbs on the prepared silica gel column. After washing with 0.1M ferrous sulfamate in 2.5M HNO₃, ²³⁹Np is eluted with 0.1M HNO₃ containing 0.02M HF. The separation of ²⁴³Am from ²³⁹Np is very effective. The purity of ²³⁹Np was found to be better than 99.5%. The proposed ²³⁹Np milking procedure is suitable for the preparation of ²³⁹Np tracer that can be used for the determination of ²³⁷Np radiochemical yield.     

Low level detection of <Superscript>135</Superscript>Cs and <Superscript>137</Superscript>Cs in environmental samples by ICP-MS

The measurement of fission product cesium isotopes ¹³⁵Cs and ¹³⁷Cs at low femtogram (fg) 10⁻¹⁵ levels in ground water by Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) is reported. To eliminate the natural barium isobaric interference on the cesium isotopes, in-line chromatographic separation of the cesium from barium was performed followed by high sensitivity ICP-MS analysis. A high efficiency desolvating nebulizer system was employed to maximize ICP-MS sensitivity ~10 cps/fg. The three sigma detection limit for ¹³⁵Cs was 2 fg/mL (0.1 μBq/mL) and for ¹³⁷Cs 0.9 fg/mL (0.0027 Bq/mL) measured from the standard with analysis time of less than 30 min/sample. Cesium detection and 135/137 isotope ratio measurement at very low femtogram levels using this method in a spiked ground water matrix is also demonstrated.     

<Superscript>239</Superscript>Np as a tracer of <Superscript>237</Superscript>Np in effluent samples and low level nuclear waste

In this paper a technique to separate and measure both isotopes (²³⁷Np and ²³⁹Np) together is presented. A combined shape pulse discrimination liquid scintillation measurement with gamma-spectrometry, permits a precise measurement after the radiochemical separation. This technique was carried out by using an Eichrom chromatographic column (TEVA) as the first step of a more complete method, applied in the Nuclear Regulatory Authority, to separate actinides in nuclear waste and liquid effluents. The MCA is 0.08 Bq/l by alpha-spectrometry and 0.22 Bq/l (2σ) by liquid scintillation counting (LSC) for 93.7% of measurement efficiency and 98.4% of chemical recovery.     

<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.     

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.     

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.     

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.     

Study on the method of preparation <Superscript>97</Superscript>Tc

Technetium-99 is one of several long-lived fission products which, when detected in the environment can give an indication of a specific nuclear activity. The most sensitive analytical technique for ⁹⁹Tc yet reported is by isotopic dilution mass spectrometry with technetium-97 as the yield tracer. A method for the preparation of ⁹⁷Tc is reported in this paper. ⁹⁷Tc was obtained by irradiation of a sample of natural ruthenium metal in a high flux reactor. After cooling for 2 years, the technetium was isolated from the sample by technique combining; deposition, solvent extraction, and ion-exchange chromatography techniques. ^99mTc and ¹⁰³Ru were used as radio-tracers for the process. The results showed that more than 70% of the Tc was recovered the decontamination factor is more than 2.3 × 10⁷. The ⁹⁷Tc was calibrated by isotope dilution mass spectrometry with ⁹⁹Tc as the yield tracer. The final yield was 29.56 μg of ⁹⁷Tc suitable for use as a mass spectrometric spike (weight % ⁹⁷Tc spike: ⁹⁷Tc, 84.77%; ⁹⁸Tc, 15.03%; ⁹⁹Tc, 0.20%).     

Production of low and high specific activity <Superscript>64</Superscript>Cu in a reactor

Production of radioactive scandium by irradiating natural titanium metal in Pakistan Research Reactor-1 was evaluated. The production rate of ⁴⁷Sc and other radioactive scandium was estimated. High specific activity ⁴⁷Sc can be produced by irradiating enriched ⁴⁷Ti in sufficient quantities needed for therapeutic applications. A new separation technique based on column chromatography was developed. Neutron irradiated titanium was dissolved in hydrofluoric acid, which was evaporated and taken in distilled water. The resulting solution was loaded on silica gel column. The radioactive scandium comes out first and the inactive titanium is removed with 2 M HCl. More than 95% radioactive scandium was recovered, while chemical impurity of titanium determined by optical emission spectroscopy was less than 0.01 μg/mL in final product.     

Hydrolytic Behavior of La<Superscript>3+</Superscript> and Sm<Superscript>3+</Superscript> at Various Temperatures

The hydrolytic species of lanthanide ions, La³⁺ and Sm³⁺, in water at I = 0.1 mol·dm^?3 KCl ionic strength and temperatures of 298.15, 310.15 and 318.15 K were investigated by potentiometry. The hydrolytic species were modeled by the HySS simulation program. From the results, the hydrolytic species of each metal ion at different temperatures were calculated using the program HYPERQUAD2013. The hydrolysis constants (log₁₀ β) of [La(OH)]²⁺ and La(OH)₃ were calculated as ?8.52 ± 0.46, ?26.84 ± 0.48, and log₁₀ β values of [Sm(OH)]²⁺, [Sm(OH)₂]⁺, Sm(OH)₃ were calculated as ?7.11 ± 0.21, ?15.84 ± 0.25 and ?23.44 ± 0.52 in aqueous media at 298.15 K, respectively. The dependence of the hydrolysis constants on the temperature allowed us to calculate the enthalpy, entropy, and Gibbs energy of hydrolysis values of each species.     

Preparation of <Superscript>166</Superscript>Dy/<Superscript>166</Superscript>Ho-chitosan as an in vivo generator for radiosynovectomy

¹⁶⁶Ho is one of the most effective radionuclides used for radiosynovectomy. One method to deliver this radioisotope to target tissue is via the ¹⁶⁶Dy/¹⁶⁶Ho in vivo generator system. The aim of this work was to prepare ¹⁶⁶Dy/¹⁶⁶Ho-chitosan (¹⁶⁶Dy/¹⁶⁶Ho-CHIT) in vivo generator for radiosynovectomy applications. ¹⁶⁶Dy obtained by the irradiation of natural ¹⁶⁴Dy target. ¹⁶⁶Dy was separated from ¹⁶⁶Ho by extraction chromatographic method (separation yield; 93% and separation factor;1.7). Chitosan labeling was performed in acetic acid with 99.3 ± 0.6% radiochemical purity. Biodistribution studies on intraarticular injected rats demonstrated high retention in the knee joint even 7 days showing no radioactivity leakage from the injection site into other organs as well as any translocation of the daughter nucleus after β^? decay of ¹⁶⁶Dy.