Experimental and FLUKA study relevant to the production of <Superscript>124</Superscript>I through <Superscript>nat</Superscript>Te(p,xn) reaction at a medium-sized cyclotron

A radiochemical methodology for the determination of ⁹⁴Nb in low-level radioactive wastes from nuclear power plant was proposed. Although ⁹⁴Nb is a strong gamma emitter, its concentration in radioactive waste samples is usually several orders of magnitude lower than that of other β–γ emitters, whose emissions interferes in the detection of the emission lines of ⁹⁴Nb. The procedure involves acid digestion, precipitation, cation exchange chromatography by using Amberlite IRA120 resin, extraction chromatography by using TEVA resin to isolate the Nb and the gamma spectrometry to its measurement. The chemical yield was 70% in average. Samples of evaporator concentrate and spent resin were analyzed. For the samples of the evaporator concentrate, the results obtained were below L _D = 9.59 × 10^?4 Bq g^?1. For the spent resin an average result of 1.54 × 10² Bq g^?1 was obtained.     

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%.     

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 of <Superscript>239</Superscript>Pu, <Superscript>144</Superscript>Ce,and <Superscript>90</Superscript>Sr Behavior During Radioactive Wastes Treatment by Thermal Plasma Technology

A computer model is used to study the volatility of some radioelements (Cerium, Plutonium and Strontium) during radioactive wastes treatment by thermal plasma technology. This model is based on the calculation of system composition using the free enthalpy minimization method, coupled with the equation of mass transfer at the reactional interface. The model enables the determination of the effects of various parameters (e.g., temperature, plasma current, and presence of oxygen in the carrier gas) on the radioelement volatility. The obtained results indicate that any increase in molten bath temperature causes an increase in the radioelement volatility. It is also found that the oxygen flux in the carrier gas strengthens the radioelement incorporation in the containment matrix. For electrolyses effects, an increase in the plasma current increases both the vaporization speed and the vaporized quantities of ²³⁹Pu, ¹⁴⁴Ce, and ⁹⁰Sr.     

Application of an experimental design to optimize a segregation method of <Superscript>129</Superscript>I and <Superscript>14</Superscript>C

In this work, the optimization of a segregation method of ¹²⁹I and ¹⁴C, two long-living radionuclides, main constituents of nuclear radioactive waste, has been developed. To be able to carry out this project, a fractional factorial experimental design was applied using 5 factors and 2 levels by factor (2^5–2). Only 8 experiments were necessary to identify the variables affecting the process, and very good recoveries of both radionuclides were obtained: (94?±?2)% for ¹²⁹I, and (99?±?1)% for ¹⁴C. The segregation of ¹²⁹I was influenced by flow (Q), volume of H₂SO₄ (V_H+), and carriers (CR), while V_H+ and time (t) played a major role in the segregation of ¹⁴C.     

Synthetic calcium aluminosilicate monolith: V. Change in the pore structure in hydration and carbonization

The physicochemical properties of synthetic calcium aluminosilicate (SCAS) monoliths produced from fly ash, limestone, and sand in a three-stage process (filtration combustion with superadiabatic heating, fine grinding, and pressing) were studied. It was found that hydration and carbonization in a SCAS monolith during long hardening under natural (laboratory) conditions lead to perfection of the structure of pores, which improves its physicochemical properties. The presence of unreacted β-Ca₂SiO₄ in the SCAS monolith throughout the hardening period ensures its high immobilizing properties under the action of the hydrosphere on the matrix containing hazardous (including radioactive) wastes because of calcium hydrosilicate gel formation, which decreases the pore space volume. Examples were given for determining the dependence of the total rate of leaching of SCAS monoliths by deionized water at 90°C on the treatment time (MCC-1 test). The rate of leaching of a SCAS-MRW monolith (where MRW is model radioactive waste of closed nuclear fuel cycle) was found to be 6.7 × 10^?7, 7.2 × 10^?7, and 8.3 × 10^?7 g cm^?2 day^?1 at MRW contents of 10, 20, and 30 wt %, respectively. The possibility of integrated solutions of some environmental problems using energy- and resource-saving technologies was considered.     

<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.     

Determination of <Superscript>93</Superscript>Zr and <Superscript>237</Superscript>Np in nuclear power plant wastes

A combined radioanalytical method for determination of ⁹³Zr and ²³⁷Np (as well as other actinoids) in radioactive wastes has been developed. Analytes were co-precipitated on iron(II)-hydroxide, separated and purified on UTEVA columns, and detected by inductively coupled plasma mass spectrometry. According to Zr and Np, 65 and 75% yields were achieved, respectively.     

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.     

Variation in structure and Li<Superscript>+</Superscript>-ion migration in argyrodite-type Li<Subscript>6</Subscript>PS<Subscript>5</Subscript>X (X = Cl,Br, I) solid electrolytes

All-solid-state rechargeable lithium-ion batteries (AS-LIBs) are attractive power sources for electrochemical applications due to their potentiality in improving safety and stability over conventional batteries with liquid electrolytes. Finding a solid electrolyte with high ionic conductivity and compatibility with other battery components is a key factor in raising the performance of AS-LIBs. In this work, we prepare argyrodite-type Li₆PS₅X (X = Cl, Br, I) using mechanical milling followed by annealing. X-ray diffraction characterization reveals the formation and growth of crystalline Li₆PS₅X in all cases. Ionic conductivity of the order of 7?×?10^?4 S cm^?1 in Li₆PS₅Cl and Li₆PS₅Br renders these phases suitable for AS-LIBs. Joint structure refinements using high-resolution neutron and laboratory X-ray diffraction provide insight into the influence of disorder on the fast ionic conductivity. Besides the disorder in the lithium distribution, it is the disorder in the S^2?/Cl^? or S^2?/Br^? distribution that we find to promote ion mobility, whereas the large I^? cannot be exchanged for S^2? and the resulting more ordered Li₆PS₅I exhibits only a moderate conductivity. Li⁺ ion migration pathways in the crystalline compounds are modelled using the bond valence approach to interpret the differences between argyrodites containing different halide ions.     

A simple method for self-attenuation correction in <Superscript>210</Superscript>Pb activity measurement in a well-type HPGe detector

A simple method for ²¹⁰Pb determination in a well-type detector for matrix with apparent densities ranging from ρ = 0.430 g cm^?3 to ρ = 2.037 g cm^?3 is presented. Ten spiked samples of ²¹⁰Pb were prepared to obtain the detector efficiency as a quadratic function of the matrix density. Then this equation was validated and successfully used to measure ²¹⁰Pb concentration activity in other samples. The equation proposed in this work is specific for each germanium detector; however it is proposed an extrapolation of the method to other well-type germanium detector by preparing a spiked sample and determining the efficiency for ²¹⁰Pb.     

Determination of <Superscript>59</Superscript>Ni in radioactive waste

Composite material PAN-DMG, containing chelating agent dimethylglyoxime (DMG) immobilized in porous matrix of binding polymer polyacrylonitrile (PAN), was used for nickel separation and concentration. Method for preparation of ⁵⁹Ni source for low energy photon spectrometry was developed using homogeneous precipitation of nickel with DMG. The proposed method was tested with two types of real radioactive waste (boric acid concentrate from nuclear power plant (NPP) evaporator and spent ion exchanger from NPP).     

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.     

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.     

Gasification characteristics to <Superscript>14</Superscript>CO<Subscript>2</Subscript> of <Superscript>14</Superscript>C radionuclide desorbed from spent resin by phosphate solutions

The removal characteristics of H¹⁴CO₃ ⁻ ions from IRN-150 mixed resin contaminated with ¹⁴C radionuclide and the gasification effects of ¹⁴C radionuclide on ¹⁴CO₂ are investigated in this study. The stripping solutions used for the removal of ¹⁴C from spent resin are NaNO₃, Na₃PO₄, NH₄H₂PO₄, and H₃PO₄. The influence of the stripping solution concentration on the desorption characteristics of an inactive HCO₃ ion into a stripping solution from IRN-150 mixed resin and the gasification of this ion to CO₂ is analyzed. The gasification behavior to CO₂ using NaOH, HNO₃, and HCl was also compared to that of phosphate solution. Spent resin stored in Wolsong nuclear power plant is used to evaluate the gasification characteristics of ¹⁴C radionuclide to ¹⁴CO₂. Gamma radionuclides such as ¹³⁷Cs and ⁶⁰Co in residual striping solutions after desorption experiments are analyzed.     

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.     

Polycyclic aromatic hydrocarbon and ionic compositions of atmospheric bulk deposition samples at a national park under the influence of intense barbecue smoke

The main aim of this study was to present the effects of barbecue smoke on a small-scale environment, a national park under the influence of intense barbecue smoke, and to scientifically support the sustainable usage of the park. Twelve-weekly bulk deposition samples were collected directly at the barbecuing area, and the samples were analysed for 16 US EPA’s priority PAH compounds and major ions. The mean concentrations of the individual PAHs in the bulk deposition samples ranged from 11.8 ng L^?1 (Ane) to 1085 ± 581 ng L^?1 (IcdP). The most frequently observed PAH compounds in the bulk deposition samples were Np, Anp, Flr, Phe, An, Flu, BkF, BaP and IcdP. The mean total PAH deposition fluxes were determined as 3.6 ± 5.6 µg m^?2 day^?1. The chloride, potassium and the sulphate fluxes were determined as 145.2 ± 267.8 µg m^?2 day^?1, 182.9 ± 291.9 µg m^?2 day^?1, and 111.9 ± 65.9 µg m^?2 day^?1, respectively. Dominant ions in the bulk deposition samples were potassium ion, chloride and sulphate which addressed as the fingerprint of barbecue grilling.     

LiFePO4/C composites from carbothermal reduction method

Using the cheap raw materials lithium carbonate, iron phosphate, and carbon, LiFePO₄/C composite can be obtained from the carbothermal reduction method. X-ray diffraction (XRD) and scanning electronic microscope (SEM) observations were used to investigate the structure and morphology of LiFePO₄/C. The LiFePO₄ particles were coated by smaller carbon particles. LiFePO₄/C obtained at 750 °C presents good electrochemical performance with an initial discharge capacity of 133 mAh/g, capacity retention of 128 mAh/g after 20 cycles, and a diffusion coefficient of lithium ions in the LiFePO₄/C of 8.80?×?10^?13 cm²/s, which is just a little lower than that of LiFePO₄/C obtained from the solid-state reaction (9.20?×?10^?13 cm²/s) by using FeC₂O₄ as a precursor.     

Industrial production and purification of <Superscript>32</Superscript>P by sulfur irradiation with partially moderated neutron fluxes and target melting

Target purification of S_α is carried out by distillation at 444±2 °C under N atmosphere and diluting the vapors in CS₂. The solution is filtered through fiberglass, Teflon and cellulose to obtain S_α by CS₂ evaporation. Once 30 g of this target are irradiated with fast neutron fluxes from 4.5 to 7.4·10¹² n·cm⁻²s⁻¹ from 6 to 12 hours, the nuclear reaction ³²S(n,p)³²P takes place. So, the irradiated S_α sample is placed in a Pyrex container situated inside a furnace as the most important piece of equipment in one aluminum and Lucite glove box. The distillation of irradiated sulfur takes place at 444±2 °C under N atmosphere during 1–2 hours. The vapors are connected to a sulfur diluter containing 20% CS₂ aqueous solution, followed by an activated carbon filter and the two similar additional sulfur diluters. Once cooled, the distillation chamber keeps the radioactive, carrier-free ³²P stuck to the wall. Then 25–50 ml of 0.1N HCl acid was injected by suction and heated again at 110±2 °C during 1 hour. The corresponding chemical reaction takes place and the labeled H₃ ³²PO₄ solution is produced. In such a way, industrial production of ³²P labeled molecules has started in Mexico, with an initial production of 3700–5550 MBq per week.     

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.