Band 107, Heft 4

Experiments were performed to evaluate production of 72 Se, parent radionuclide of the positron emitter 72 As, at high energy at the Brookhaven Linac Isotope Producer (BLIP). Excitation functions for 75 As(p, xn) 72/75 Se in the 52-105 MeV energy range were measured by irradiating thin gallium arsenide (GaAs) wafers. Maximum cross section value for the nat As(p, 4n) 72 Se reaction in the energy range was 103±9 mb at 52±1 MeV. Production size GaAs and arsenic metal (As°) targets were irradiated with 136 μA and 165 μA beam current possessing an initial Linac energy of 117 MeV. A total of 3.77±0.1 GBq (102±3 mCi) of 72 Se was produced from a GaAs target at a calculated target entrance energy of 105.4 MeV, and 13.8±0.3 GBq (373±8 mCi) of 72 Se from an As° target at a calculated incident energy of 49.5 MeV irradiated for 116.5 h and 68.9 h, respectively.

This paper is an attempt to find out thorium oxide dissolution mechanism in HNO 3 -HF mixture. In a previous paper, several parameters effects on thorium oxide dissolution have been described, with specific focus on hydrofluoric acid effect, which can lead to an increase of the dissolution rate if present in small amount, but precipitates as ThF 4 at higher content. Based on this previous study, experimental data were fitted using several dissolution models in order to find out the best one. Finally, a revisited model based on literature and considering the ThF 4 formation was proposed. It describes the main steps of dissolution and is able to fit the experimental data for a wide range of solution compositions. This point is crucial since it allows considering an extrapolation of the established model to not-yet-studied conditions.

The three-dimensional (3D) carbonaceous nanofiber and Ni-Al layered double hydroxide (CNF/LDH) nanocomposite was successfully prepared by a facile one-step hydrothermal methodology. Characterization of scanning electron microscope (SEM), transmission electron microscopy (TEM), X-ray photoelectron spectroscopy (XPS), XRD, and Fourier transformed infrared spectroscopy (FTIR) provided a demonstration that the modified CNF/LDH nanocomposite possessed abundant functional groups, for instance, metal-oxygen surface bonding sites (Ni–O as well as Al–O) and free-metal surface bonding sites (C–O, C–O–C, as well as O–C=O). The elimination of representative radionuclide (i.e. U(VI)) on the CNF/LDH nanocomposite from aqueous solutions was explored as a key function of pH, ionic strength, contact time, reaction temperature as well as radionuclide preliminary concentrations with the use of the batch methodology. As revealed by the findings, the sorption of radionuclides on CNF/LDH nanocomposite adhered to the pseudo-second-order kinetic model as well as Langmuir model. The maximum elimination capacity of U(VI) amounted to be 0.7 mmol/g. The independent of ionic strength shed light on the fact that inner-sphere surface complexation mainly overpowered radionuclide uptake by the CNF/LDH nanocomposite, which was further verified through the combination of FTIR and XPS spectral analyses. The abovementioned analyses shed light on the fact that the CNF/LDH nanocomposite can be regarded as a latent material to preconcentration radionuclides for environmental remediation.

A series of uranyl sulfates and selenates templated by protonated forms of amino acids (glycine, α- and β-alanine, threonine, nicotinic, and isonicotinic acid) has been prepared via isothermal evaporation of strongly acidic solutions. Their structures have been refined by the direct methods and can be classified as inorganic [(UO 2 ) m ( T O 4 ) n (H 2 O) k ] ( T =S 6+ , Se 6+ ) moieties combined with the protonated amino acid cations, water molecules and hydronium ions. Their overall motifs demonstrate common features with related structures templated by organic amines. The role of carboxylic acid groups depends on the nature of the corresponding amino acid. They can either link two protonated organic moieties into dimers, or contribute to hydrogen bonding between organic and inorganic parts of the structure. The ammonium ends of the amino acid cations form strong directional bonds to the oxygens of the uranyl and T O 4 anions.

A novel and useful extraction spectrophotometric method for the rapid determination of uranium and free acidity in nuclear fuel process samples using the reagent, 2,3-dihydroxynaphthalene as an extractant as well as color forming agent is described. Uranium at milligram level forms a yellowish- orange anionic complex with the ligand, 2,3-dihydroxynaphthalene over a pH range, 10–12. This anionic complex is easily extracted into ethylacetate as an ion-pair complex with cetyltrimethylammonium cation (CTA + ). The absorbance of the ion-pair complex in ethylacetate is measured at 390 nm. The molar absorptivity and Sandell’s sensitivity of the system being 5.0×10 3 L·mol −1 cm −1 and 0.047 μg cm −2 , respectively at 390 nm. This very reagent (2,3-dihydroxynaphthalene) also forms a yellowish-orange nitro product with free nitric acid (HNO 3 ) in the presence of concentrated sulfuric acid. This nitro- product is easily extracted into the same solvent (ethylacetate) and absorbs maximum at 380 nm. The molar absorptivity (ε) and Sandell’s sensitivity of the system being 2.4×10 3 L·mol −1 cm −1 and 0.027 μg cm −2 , respectively. By using the single chromophoric agent (2,3-dihydroxynaphthalene), both uranium and free acidity can be easily and reliably determined separately in the process solution. The relative standard deviations (RSD) are in the range, 0.5–2.0% and 1.5–2.0%, respectively for uranium and free nitric acid determinations. The method has been applied to nuclear fuel process solutions, and the results obtained have been found to be favorably comparable with those obtained from standard methods.

Recently, technologists try to develop novel gamma radiation shielding materials instead of traditional materials such as lead and concrete with improved performance in gamma radiation shielding in medical applications and nuclear reactors. For this purpose, alloys such as stainless steel (SS) and carbon steel (CS) attracted much attention, these days. Preliminary results on such alloys have shown better attenuation of γ rays as compared to traditional shielding materials. This work aimed to conduct research on different alloy samples to evaluate their radiation attenuation efficiency and their suitability for radiation shielding when utilized in nuclear facilities. The mass attenuation coefficients for eight alloy samples were measured at different photon energies ranging from 80.997 to 1332.501 keV using transmission geometry. From the mass attenuation coefficients, different photon attenuation parameters such as half value layer, mean free path, effective atomic number, and radiation protection efficiency were evaluated. In addition, the equivalent atomic number and the exposure buildup factor were calculated using G-P fitting method for photon energy ranging from 0.015 MeV to 15 MeV at different penetration depth. The results showed that the Z eff values remain almost constant for all samples except W72/Cu28 in which the Z eff for this sample tends to decrease with the energy. The lowest value of half value layer is found for the alloy sample Ta97.5/W2.5 and the highest value is found for the alloy sample In50/Sn50. The Ta97.5/W2.5, Ta90/W10, Ta95/W5 samples demonstrated good radiation attenuation properties.

The mass attenuation coefficients (μ/ρ) of calcium tungstate, ammonium tungsten oxide, bismuth tungsten oxide, lithium tungstate, cadmium tungstate, magnesium tungstate, strontium tungsten oxide and sodium dodecatungstophosphate hydrate were measured at 14 photon energies in the energy range of 81–1333 keV using 22 Na, 54 Mn, 57 Co, 60 Co, 133 Ba and 137 Cs radioactive sources. The measured μ/ρ values were compared with those obtained from WinXCOM program and the differences between the experimental and theoretical values were very small. The bismuth tungsten oxide has the highest μ/ρ among the present samples in the studied energy region. From the μ/ρ values, we calculated the half value layer, tenth value layer and mean free path, and the results showed that ammonium tungsten oxide (which has the lowest density) and bismuth tungsten oxide (which has the highest density) possess the highest and lowest values of these three parameters, respectively. Additionally, from the incident and transmitted photon intensities, we calculated the radiation protection efficiency (RPE). The bismuth tungsten oxide was found to have RPE 98.53 % at 81 keV, which has the maximum value among the present samples and this suggested that bismuth tungsten oxide is the best to be chosen as the γ radiation shielding material candidate among the selected samples.

In the present work, we developed three new high temperature resistant heavy concretes as novel radiation shielding materials. For this purpose, chrome ore (FeCr 2 O 4 ), hematite (Fe 2 O 3 ), titanium oxide (TiO 2 ), aluminum oxide (Al 2 O 3 ), limonite [FeO (OH) nH 2 O], siderite (FeCO 3 ), barite (BaSO 4 ), nickel oxide (NiO) materials and alumina cement were used. Mass combination ratios of components and total macroscopic cross sections (scattering, absorption, capture, fission) of the samples were calculated by using GEANT4 code. The resistances of the prepared samples were evaluated in terms of compression strength after exposure at the 1000 °C temperature. Neutron equivalent dose rate measurements were carried out by using 4.5 MeV 241 Am-Be neutron source and BF 3 detector. All results were compared with normal weight concrete and paraffin. The results of neutron dose indicate that neutron absorption ability of the new heavy concretes is higher than normal weight concrete and paraffin. In addition to neutron measurements, different γ-ray shielding parameters such as mass attenuation coefficient (MAC), effective atom numbers (Z eff ), half value layer (HVL) and mean free path (MFP) have been calculated using WinXCOM software in order to investigate the effectiveness of using the prepared concretes as a radiation shielding materials. Gamma-ray results were compared with concretes and Pb-based glass.