Band 106, Heft 2

Excitation functions of the nat Ni(α,x) 60,61 Cu and 64 Ni(α,p) 67 Cu reactions were measured using the stacked-foil activation technique. The experimental data achieved were compared with literature data as well as with nuclear model calculations performed using the code TALYS-1.8. Integral yields from the respective thresholds to 44 MeV were deduced from the measured excitation curves. The >99% pure 61 Cu can be produced using the energy window of E α =20→7 MeV, the yield amounting to 116 MBq/μAh. After a 2 h cooling time, the short-lived 60 Cu (T 1/2 =23 min) impurity will be reduced to <0.1%. Due to low isotopic abundance of 64 Ni, the enriched 64 Ni target would be needed for the production of 67 Cu via the (α,p) reaction. The cost would, however, be very high. In a few cases, particularly above 24 MeV, we have given new data points. In general, our measurements have strengthened the database.

The dissolution of uranium or uranium-plutonium carbide fuel in nitric acid leads to ~50% carbon evolved as carbon dioxide, the remainder remains in the solution as soluble organics. These dissolved organic molecules interfere with the solvent extraction of uranium and plutonium by complexing to the actinide ions and decreasing the efficiency of their extraction. Experiments reported here describe two series of experiments assessing the uranium carbide dissolution liquor treatment by prolonged boiling and electrochemical oxidation. Plutonium losses to aqueous and solvent raffinates are observed for untreated liquors, highlighting that mineralisation of dissolved organics is necessary to reduce the complexing effects of organic acids to an extent that permit efficient operation of a solvent extraction process both in the first solvent use (considered here) and for maintaining solvent quality during industrial solvent reuse in the highly active cycle. Solution carbon analysis and 30% TBP solvent extraction batch tests of uranium and plutonium originating from dissolved uranium carbide liquors untreated and after treatment are compared. These experiments demonstrate the reprocessing of uranium carbides by direct dissolution coupled to a mineralisation process, can achieve near quantitative uranium and high plutonium recoveries (99.9%).

A combination of neutral and acidic extractant has been proposed for the single-cycle separation of trivalent actinides from high-level liquid waste (HLLW). The nature of acidic extractant in the combined solvent formulation plays a profound role in deciding the extraction and stripping of trivalent actinides. Therefore, the extraction behavior of Am(III) in a solution of tetra-bis(2-ethylhexyl)-diglycolamide (TEHDGA) and acidic extractant (HA) was studied from nitric acid medium. The acidic extractants chosen were bis(2-ethylhexyl)phosphoric acid (HDEHP), bis(2-ethylhexyl)phosphonic acid (PC88A) and bis(2,4,4-trimethylpentyl)phosphinic acid (CYANEX-272) whose pK a values were 3.24, 4.51 and 6.37, respectively. The distribution ratio of Am(III) was measured as a function of various parameters such as concentration of nitric acid, TEHDGA, HA etc. The data were compared with those obtained in individual solvent systems namely 0.1 M TEHDGA/ n -DD and HA/ n -DD. Slope analysis of the extraction data indicated the synergic participation of both TEHDGA and HDEHP in the extraction of Am(III) at all acidities. However, antagonistic effect was observed at lower acidity when TEHDGA was mixed to PC88A or CYANEX-272 present in n -DD. Accordingly, a suitable mechanism has been proposed for the extraction of Am(III) at all acidities using these combined solvent formulation. Studies with fast reactor simulated high level liquid waste indicated that extraction of Am(III) was accompanied by co-extraction of lanthanides and unwanted metal ions such as Zr(IV), Mo(VI), Y(III) and Pd(II). However, addition of trans-1,2-diaminocyclohexane-N,N,N,N′-tetraaceticacid (CyDTA) reduced the extraction of unwanted metal ions. Batch extraction and stripping studies indicated the possibility of using 0.1 M TEHDGA+0.25 M HDEHP in n -dodecane for the single cycle separation of Am(III) from FR-SHLLW.

Plutonium dioxide samples have been studied by static leach tests at a temperature of 25 and 90°C. A normalized Pu mass loss was observed from polycrystalline (without correction for matrix porosity) and monocrystalline 239 PuO 2 samples occurring at a rate of 10 −4 –10 −3 g/m 2 in distilled water. This is comparable with some ceramic waste forms proposed for Pu immobilization. In contrast, the Pu release rate for polycrystalline 239 PuO 2 samples doped with 238 Pu is significantly higher for distilled water and acidic solution, ranging from 10 −2 to 10 −1 g/m 2 .

This article summarizes the results obtained studying the interaction between polonium and stainless steel 316L in different atmospheres by the thermochromatography method. This issue is particularly important in terms of licensing of the MYRRHA prototype ADS reactor, where highly radiotoxic α-emitting polonium isotopes are produced in lead-bismuth eutectic used as both spallation target material and coolant, while stainless steel 316L is going to be the main structural material. The polonium adsorption enthalpy on stainless steel 316L in inert atmosphere was measured as −147±6 kJ mol −1 . The findings obtained using reducing atmospheres show that the affinity of polonium for stainless steel 316L increases with the metallic character of the stainless steel surface. The presence of moisture in inert and reducing atmosphere does not significantly influence the polonium adsorption. Oxidizing atmospheres induce the formation of polonium dioxide which has a stronger interaction with the stainless steel 316L, with an adsorption enthalpy of −225±6 kJ mol −1 . A prolonged exposure of the stainless steel to strong oxidants, such as pure oxygen, reduces the metallic character of the stainless steel surface and, consequently, its affinity to the polonium oxide. Furthermore, it was found that in all experiments at least 5 ppm of the total polonium present in the starting material were not retained on the stainless steel 316L surface at all.

A method for obtaining 156, 155, 154m2, 154, 153 Tb radiotracers by the irradiation of a europium oxide target of natural isotopic composition by 27 MeV α-particles is proposed. Terbium can be efficiently separated from bulk of europium by the reduction of the latter by zinc in an acidic solution and precipitation as EuSO 4 . The optimum Zn/Eu 3+ and (NH 4 ) 2 SO 4 /Eu 3+ molar ratios are 20 and 3, respectively. Terbium is additionally purified from europium and gadolinium by extraction chromatography using LN Resin. It is demonstrated that optimum separation is attained in 0.6 M HNO 3 . The Tb/Eu separation coefficient was ~5·10 5 . The yield of terbium was about 90%. Time of all steps was 1.5–2 h. The proposed procedure makes it possible to obtain no carrier added terbium radiotracers.

The essential trace element selenium is a focus of attention due to its effects on human health, there being consequences of both its deficiency and excess. Due to the ultra-trace content of selenium, the neutron activation analysis method (NAA) is difficult to apply. We therefore made use of the radiochemical neutron activation analysis (RNAA) to determine Se at low level concentrations in several consumed food items in Algeria. A radiochemical procedure based on liquid–liquid separation was established in our laboratory. In this research we focused on the determination of selenium in two species of coffee: Arabica and Robusta. The accuracy of the method was assessed by analyzing the certified reference material NIST-SRM 1573a (tomato leaves). The results obtained show a selenium variation from 0.025 to 0.052 μg/g in coffee beans and an average yield of the separation of about 85%. The results of this study were compared with those obtained with samples from Brazilian, Caribbean, Indian and Kenyan coffee beans.

Batch sorption and column experiments were conducted to investigate and compare sorption and transport behavior of 90 Sr on the assumption of seawater intrusion at nuclear power plants. Batch sorption experiments were carried out on fractured rocks and bedrocks using synthetic groundwater and seawater both spiked with 90 Sr. In general, higher 90 Sr sorption occurred on fractured rock samples than on bedrocks, because of the presence of weathered secondary minerals (iron oxide and clay) on fractured rock surfaces. However, one particular bedrock sample (WSP-B) which has higher porosity and carbon amount than fractured rock samples also showed the higher 90 Sr sorption than its comparable fractured rocks. For all batch sorption studies, 90 Sr sorption distribution coefficient, K d decreased from groundwater to seawater environment due to the higher ionic strength of seawater (6.4×10 −1 –7.7×10 −1 M) compared to groundwater (4.0×10 −3 –6.0×10 −3 M). The three different ionic strength solutions were used in column experiments, and the results showed that transport behavior of Sr through a fractured rock had similar sorption trend to batch sorption results. The highest mobility (or least retardation) for Sr was found for 100% seawater solution compared to the highest retardation (or least mobility) for 100% groundwater solution. These sorption and transport data of Sr on solid materials contacted with various ionic strength solutions corroborate empirically defensible information for assessment of radioactive contamination in groundwater below the NPP sites located nearby shores. In addition, the experimental data will be incorporated to improve transport models of 90 Sr in the subsurface environment for severe nuclear accidents.

In this study, a detailed γ-ray spectrometer survey was performed to obtain the baseline data on natural and fallout radioactivity in soil samples (10–30 cm depth) collected from Iğdır which is a province in Eastern Anatolia Region of Turkey and located along the border with Armenia. Metsamor Nuclear Power Plant in Eastern Armenia is only about 20 km away from Iğdır province. The average activity concentrations of 226 Ra, 232 Th, 40 K and 137 Cs in the soil samples were found as 19.1±0.6 (range: 12.8–31.2 Bq kg −1 ), 21.9±0.6 (range: 11.6–36.8 Bq kg −1 ), 437.7±10.3 (range: 285.0–613.7 Bq kg −1 ) and 11.8±1.1 Bq kg −1 , (range: 2.5–51.1 Bq kg −1 ), respectively. The radiological hazards caused by ionizing radiation emitted from these radionuclides on the population living in the region were assessed by estimating radiological parameters such as absorbed γ dose rate in outdoor and the corresponding annual effective dose rate from external exposure, annual effective dose rate from inhalation of radon and excess lifetime cancer risk. The results revealed that there is no significant radiological hazard for human population because the average values of radiological parameters are within the recommended limits.