Volume 103, Issue 2

The cross sections of the reaction 98 Mo( n ,  γ ) 99 Mo at neutron energies of 0.0334 eV and 0.0536 eV were measured for the first time using the activation technique, and the resulting values of 116 ± 7 mb and 91 ± 5 mb, respectively, are in good agreement with the JENDL-4 and ENDF/B-VII data files. Those measured values were extrapolated to the energy 0.0253 eV assuming 1/v dependence in the thermal region and the results were compared with the integrally measured values reported in the literature. The new data are thus useful to check the evaluated excitation function.

A novel 2-oxy-N,N-dioctylacetamide anchored on Merrifield resin was synthesized and characterized by elemental analysis, TG-DTA and IR spectroscopy, for the first time. The extraction behavior of Am(III) and Eu(III) from nitric acid medium in the dioctylacetamide anchored resin (R-DOA) was studied as a function of various parameters such as duration of equilibration, concentration of nitric acid and sodium nitrate etc. The distribution coefficient ( K d ) of Am(III) and Eu(III) in R-DOA increased with increase in concentration of nitric acid. The K d values reached a maximum at 3–4 M nitric acid concentration in both cases, followed by decrease in K d values. Rapid extraction of Am(III) and Eu(III) from nitric acid medium was observed in the initial stages of equilibration followed by the establishment of equilibrium occurred within 60 min. The extraction data was fitted in to the first order rate equation. The recovery of Am(III) and Eu(III) from the loaded organic phase was studied using dilute nitric acid. The results revealed that 2-oxyacetamide functionalized resin is the promising candidate for the separation of trivalent actinides and lanthanides from nitric acid medium.

As part of the studies on the development of alternate extractants for fast reactor fuel reprocessing, a mixer-settler facility for performing counter-current solvent extraction runs with radioactive materials has been set up and commissioned in our laboratory for flow sheet development experiments. Mixer-settler runs were carried out with 1.1 M TiAP/Heavy Normal Paraffin (HNP) as the solvent using the above facility to test the equipment as well as the behaviour of the solvent for continuous solvent extraction processes. Stage profile data generated for the extraction and stripping of nitric acid with TiAP based solvent have been reported. Studies carried out on the continuous counter-current extraction and stripping of U(VI) for 1.1 M TiAP/HNP-U(VI)-HNO 3 system under high solvent loading conditions are also reported in this paper. Overall and stage-wise mass balance data for the above runs are also discussed.

Different routes can be used to produce the PET isotope 68 Ga by proton reactions on different targets. The experimental excitation functions for the reactions 68 Zn( p ,  n ) and 72 Ge( p ,  nα ) were collected from previous studies and compared with each other, for each reaction, in order to get the most trusted data. Some data were excluded because they gave either too high or too low cross section values. The others were averaged together to give a kind of trusted data set for the production of this important radioisotope. The production of the radionuclide 68 Ge, used in the generator 68 Ge/ 68 Ga, through proton reactions was also discussed in the same manner. Model calculations of excitation functions of the above mentioned reactions were performed, using both TALYS-1.4 and EMPIRE-3.1 codes, and the results were compared with the averaged experimental data. Some statistical parameters were introduced to control the quality of the overlapping between experimental data and model calculations. The integrated yields for the different reactions leading to 68 Ga were estimated. The various routes were discussed on the basis of the given results.

Germanium-68 (Ge-68) is produced by proton irradiation of a gallium metal target and purified by organic extraction. The Ge-68 can be used in a medical isotope generator to produce Gallium-68 (Ga-68) which can be used to radiolabel PET imaging agents. The emerging use of Ge-68 in the Ga-68 medical isotope generator has caused us to develop a new purification method for Ge-68 that does not use toxic solvents. The purpose of this work was to develop a production scale separation of Ge-68 that utilizes a leaching step to remove a bulk of the gallium metal, followed by purification with Sephadex © G25. Production scale (300 mCi) purification was performed with the new method. The purified Ge-68 contained the highest radioactivity concentration of Ge-68 produced at BNL; the sample meet Department of Energy specifications and the method had an excellent recovery of Ge-68.

The synthesis, characterization and biological evaluation of five neutral and lipophilic 99m Tc-complexes, so-called cytectrenes, obtained from N -substitutedferrocenecarboxamide derivatives are reported. N -substitutedferrocenecarboxamide starting materials were obtained in two steps, with good yield and were fully characterized by classical spectroscopic methods including X-ray diffraction analysis for one of them. Using a microwave strategy for the 99m Tc-radiolabelling step, each cytectrene were obtained quickly (radiolabelling time < 5 min), from modest to good yield. The 99m Tc-complexes, characterized by HPLC comparison with cold rhenium complex analogues, are stable, neutral and lipophilic (log  P o/w ranged between 1.8 and 2.9). Unfortunately, despite such suitable features, in vivo studies of two of them gave poor results, in terms of brain uptake. Both radiocompounds exhibited the maximum brain accumulation of 0.31% ID/g and 0.26% ID/g at 5 min post-injection, respectively, followed by a very fast washout from the brain (0.06% ID/g and 0.07% ID/g at 30 min post-injection, respectively). Although our ligand systems exhibited high stability against exchange reactions with blood proteins, the high radioactivity level in stomach, increasing with time, suggests in vivo decomposition of our complex to pertechnetate.

The high concentrations of opiate and solvent in wastewater are toxic to biological life and affect the aquatic environment. Therefore, it must be treated by an advanced treatment process such as ionizing radiation. Effect of organic media on morphine and codeine during gamma irradiation was determined for the first time in this paper. Samples were irradiated at ambient temperature and in air environment at various doses (0, 10, 20, 30, 40, 50 and 60 kGy). Gamma irradiation-induced changes in the molecular structure of morphine and codeine were monitored by direct infusion electrospray ionization mass spectrometry in positive ion mode. The mass of the by-products were appeared to be more than the mass of the original alkaloids. Molecular structures of the by-products and reaction pathways were proposed. Oxygenated morphine and oxygenated codeine were identified in the presence of oxygen. However, solvent radical addition reactions were observed as the main mechanism for the by-product formation in oxygen-free irradiation. The results indicated that 89% of morphine and 98% of codeine were degraded at dose of 50 kGy. In addition, alkaloids and their by-products were not detected above 50 kGy. Here, we demonstrated that ionizing radiation process is a promising method to remove morphine and codeine in solvent/opiate rich industrial wastewater.

In this work characteristics of the polycarbonate films with 20 μm in thickness containing different weight percentage of Benzo- δ -sultam were studied for use as a high dose dosimetry system in radiation processing facilities. The sensitivity of the dosimeters and the linearity of dose-response curves were investigated under 60 Co γ -rays in a dose range of 0–100 kGy, and obtained results were compared with the commercial CTA and FWT film dosimeters. The results show that the absorbance at 348 nm depends linearly on the dose in the investigated dose range. The effects of pre-irradiation (shelf-life) and post-irradiation storage in dark and in indirect sunlight are also discussed. The results show that the dosimeters characteristics are stable within 1% at 25 ℃, 3 months after the irradiation.

An optimization of concrete used for immobilization of radionuclides 60 Co and 137 Cs, is presented. A relatively simple mathematical model is given, which permits minimization of leach rate and permeability and maximization of compressive strength. An optimal solution, based on experimental data, is given. These results will be used for a future Serbian radioactive waste disposal center.