Band 97, Heft 9

Excitation functions of the reactions 89 Y( p , n ) 89 Zr, 89 Y( p ,2 n ) 88 Zr and 89 Y( p , pn ) 88 Y were measured from their respective thresholds up to 17.7 MeV, with particular emphasis on data for the production of the medically important radionuclide 89 Zr. The conventional stacked-foil technique was used, and the samples for irradiation were prepared by the sedimentation process. The excitation functions obtained were compared with those calculated theoretically using the ALICE-IPPE code. The measured data were also compared with the available literature values. From the experimental data the theoretical yields of the investigated radionuclides were calculated as a function of proton energy. The excitation function for the 89 Y( p , n ) 89 Zr reaction measured in the proton energy range 5.1 to 17.7 MeV shows a pronounced broad peak in the energy range 16→11 MeV. On the other hand and because of the high threshold energy of the two other reactions 89 Y( p ,2 n ) 88 Zr and 89 Y( p , pn ) 88 Y, their excitation functions were measured only in the proton energy range 14.3 to 17.7 MeV. Differential and integral yields for the production of the three radionuclides were calculated. It was found that the suitable energy range for the production of 89 Zr is 14→9 MeV; over this energy range the 89 Zr amounts to 58 MBq/μA h. In the proton energy range provided by MGC-20 cyclotron [ E p (max)=18 MeV], only 89 Zr can be produced in a suitable activity and with a high purity. Production of 88 Zr and 88 Y needs proton energy higher than that provided by this type of cyclotron.

The primary patterns of isolation of actinides from their dioxides with supercritical carbon dioxide containing adducts of tri- n -butyl phosphate and other organic reagents with nitric acid has been investigated. The effective extraction of actinides from their dioxides was achieved using these adducts under common (ambient) conditions without supercritical carbon dioxide, which makes the performance of the process significantly easier. Extraction of uranium and the primary fission products from a simulated spent nuclear fuel has been studied. It was shown that at almost quantitative extraction of U (>99%), most 95 Zr and 95 Nb (up to 90%) as well as 131 I, partially 103 Ru (∼30%), 99 Tc (20%) and lanthanides (up to 10%) also passed into the adduct phase. In this case complete separation of U from 137 Cs, 85 Sr, 140 Ba as well as from the bulk of 140 La, 141–144 Ce and 147 Nd was observed. After dissolution of the adduct containing U and partially fission products in liquid carbon dioxide the extract was transported into a separation column of a planetary centrifuge, and separation of U from remaining fission products was performed by countercurrent chromatography in the dynamic mode.

A number of cross-sensitive chemical sensors for determination of stable fission products (rare earth metals) were developed using different extracting agents that are known for their applicability in the reprocessing cycle of spent nuclear fuel. A novel “electronic tongue” multisensor system was suggested comprising an array of such sensors. The developed “electronic tongue” was tested in solutions simulating real reprocessing media and appeared being capable of simultaneous quantitative determination of several rare earth metals — typical constituents of spent nuclear fuel, e.g. cerium, neodymium and samarium. These results show a good promise for the development of a new analytical approach to multicomponent analysis in nuclear reprocessing industry.

The influence of both hydrogen peroxide (H 2 O 2 ) and bicarbonate (HCO 3 - ) on the dissolution of UO 2 has been studied in this work. Two different series of experiments have been carried out using a flow-through reactor. In the first series, the influence of H 2 O 2 concentration (between 10 -6 and 5×10 -4  mol dm -3 ) on the dissolution rate of UO 2 has been studied at a fixed bicarbonate concentration of 2×10 -3  mol dm -3 . An increase in the dissolution rate is observed as the concentration of hydrogen peroxide increases. In the second series, the influence of bicarbonate (between 10 -4 and 10 -2  mol dm -3 ) on the dissolution rate of UO 2 has been studied in the presence of a fixed hydrogen peroxide concentration (10 -4  mol dm -3 ). The main result was that UO 2 dissolution rates increased with bicarbonate concentration. From the experimental data, an oxidative dissolution model has been developed that can reproduce spent nuclear fuel dissolution rates obtained under relatively low oxygen concentrations. Under these conditions, the influence of radiolysis products, rather than O 2 concentration, is expected to determine the oxidative dissolution rates of the fuel.

The dissolution behaviour of powdered commercial spent fuel (UO 2 with burn-up of 53 MW/d kg U) has been studied in a carbonate-containing solution ([HCO 3 - ] =0.001 mol dm -3 ) by using a flow-through reactor specially designed for the use in a hot cell. This method allows studying spent fuel dissolution while avoiding the parallel process of secondary solid phase formation. The dissolution behaviour of U, Np, Pu, Sr and Cs was studied. The main trend of the results obtained in this work is that only neptunium releases congruently with uranium (FIAP Np /FIAP U =1.21±0.01) because both strontium and caesium have higher FIAP values (FIAP Sr /FIAP U =2.3±0.8; FIAP Cs /FIAP U =5±1) and plutonium lower (FIAP Pu /FIAP U =0.07±0.02). The FIAP value for uranium at the steady-state is 4(±2)×10 -4 .

The radiolytic degradation of the 6,6′-bis(5,5,8,8tetramethyl-5,6,7,8-tetrahydro-benzo[1,2,4]triazin-3-yl)-[2,2′]bipyridine (CyMe 4 BTBP) based SANEX (selective actinide extraction) solvent has been investigated. As the solvent used in the extraction process is designed to separate trivalent actinides from lanthanides, the radiolytic degradation is mainly due to alpha decay of extracted minor actinide isotopes. A calculation of dose-rates was done by estimating the concentration of minor actinides in the solvent by fuel burn-up calculations and assumptions on dilutions in the subsequent reprocessing steps. The calculations showed that the main isotopes responsible for the dose-rate are 242 Cm, 244 Cm and 241 Am. 242 Cm is short-lived and has an impact only at short cooling times before reprocessing of the spent fuel. The dose-rates to a SANEX solvent in the reprocessing of standard spent LWR fuels are burn-up dependent and range from at least 0.03–0.2 kGy/h for UO 2 fuels and from 0.4 to 0.8 kGy/h for MOX fuels. Fast reactor fuels yield dose-rates over 1 kGy/h. Based on these results, several radiolysis experiments were carried out in order to compare the effect of low LET external gamma radiation (0.2 kGy/h) and internal alpha radiation with different dose-rates (0.05, 0.2 and 1.0 kGy/h). Significant radiolytic degradation was shown in the gamma radiolysis and in the alpha radiolysis experiment at a dose-rate of 1 kGy/h. These experiments were continued up to an absorbed dose ∼1200 kGy and >300 kGy, respectively. Comparing the alpha radiolysis results for 0.2 kGy/h and 1.0 kGy/h, up to an absorbed dose of ∼120 kGy, no significant difference in the degradation for the different dose rates could be seen. The radiolytic degradation rate for gamma radiation was 40% higher than for alpha radiation.

Arsenic-77 ( T 1/2 =1.6 d) was produced by irradiating natural germanium in Pakistan Research Reactor-1. The nuclear reaction 76 Ge( n ,γ) produces 77 Ge, which decays by emission of β - particles into 77 As. The neutron irradiated target was dissolved in aqua regia, excess of acid was removed by evaporation and finally the solution in basic media was passed through hydrous zirconium oxide (HZO) column. The Ge was quantitatively retained on HZO, while 77 As was present in the effluent. More than 90% 77 As was recovered. The chemical impurity of Ge in 77 As was <0.01 μg/mL.

fac (S)-[Ir(aet) 3 ] (aet=2-aminoethanethiolate) is N3S3 metalloligand, which forms S-bridged polynuclear complexes with transition metal ions. A complex analogous to [Re{Ir(aet 3 )} 2 ] 3+ was formed by the reaction of 99m TcO 4 Na and fac (S)-[Ir(aet) 3 ] in the presence of SnCl 2 2H 2 O. A simple method for radiollabelling of fac (S)-[Ir(aet) 3 ] with 99m Tc was developed with radiolabelling efficiency higher than 99%. Effects of SnCl 2 ·2H 2 O concentration, time and pH on the radiolabelling efficiency were determined and compared with 99m Tc-[Rh(aet) 3 ]. 99m Tc-[Ir(aet) 3 ] was also characterized by electrophoresis, HPLC, biodistribution studies in rats and scintigraphy in rabbit. Higher uptake by kidneys showed rapid distribution of the labelled fac (S)-[Ir(aet) 3 ]. Scintigrams show that liver, bladder and skeletal uptake were significant. The results of biodistribution showed that 99m Tc-[Ir(aet) 3 ] uptake is quite rapid as compared to 99m Tc-[Rh(aet) 3 ].

With the aim of classifying prehispanic pottery from Cerro de Las Ventanas site, Juchipila, Zacatecas, México, instrumental neutron activation analysis (INAA) was used to analyze ceramic samples at the University of Missouri Research Reactor Center. Thirty-two chemical elements were measured: Al, As, Ba, Ca, Ce, Co, Cr, Cs, Dy, Eu, Fe, Hf, K, La, Lu, Mn, Na, Nd, Rb, Sb, Sc, Sm, Sr, Ta, Tb, Ti, Th, U, V, Yb, Zn, and Zr. Two multivariate statistical methods, cluster analysis and principal component analysis, were performed on the dataset to examine similarities between samples and to establish compositional groups. The statistical analyses of the dataset suggest that the pottery samples form a unique chemically homogeneous group, with the exception of one pottery sample. The compositional data were compared to an existing Mesoamerican ceramic database. It was found that the newly generated data fit best with data from a previous chemical analysis of pottery from the Malpaso Valley. However, despite the apparent similarity, pottery samples from the site of Cerro de Las Ventanas represent a new and unique chemical fingerprint in the region.

Radiochemistry deals with the chemistry of the radioactive elements. In the nuclear industry successful fuel reprocessing, high-level waste treatment, and long-term storage of spent fuel depend on an understanding of the radiochemistry of actinides and fission products in these settings. Radiation chemistry is concerned with the chemical effects of ionizing radiation, with the most common types of radiation encountered by the radiochemist being low linear energy transfer (LET) β - and γ radiation, and higher LET α radiation. These radiations can have profound and important effects on radiochemistry, including changes in metal oxidation states and degradation of the organic ligands designed to complex radioelements. This may occur by direct action of the incident radiation on compounds present with high abundance or by reaction with radiolytically produced reactive species for trace components, such as the complexing agents. This review examines the role of reactive species created in irradiated aqueous and organic solution and their effects on radiochemistry. The sources and nature of these reactive species are discussed. Examples of radiation chemical effects are provided related to solvent extraction of the actinides from acidic solution, metal complexation and technetium redox chemistry in alkaline tank waste, and the corrosion of spent fuel stored in repository brine.