Band 90, Heft 2 -

Employing the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several ( n, 2n ), ( n, p ) and ( n , α) reactions on isotopes of europium. Our measurements constitute the first systematic studies. Of special interest was the investigation of 153 Eu( n, p ) 153 Sm reaction for the production of no-carrier-added 153 Sm in a nuclear reactor. Using 100% enriched 153 Eu target, 97.21 MBq 153 Sm per batch can be produced which is, however, not sufficient for medical application.

Charged particle activation of Al-foil with 46 MeV 7 Li 3+ beam results in the formation of no-carrier-added radionuclides 24 Na and 28 Mg in the matrix. Two popular extractants, namely, 18-Crown-6 ether and di-(2-ethylhexyl)phosphoric acids were used to separate no-carrier-added radionuclides. It was found that HDEHP is more efficient for separation of these two adjacent s-block elements in their no-carrier-added state.

The kinetics of the oxidation of plutonium(IV) by manganese dioxide were studied in 1.0 M NaCl over the pH range from 2.5 to 8.2 with variable concentrations of manganese dioxide from 0.01 m 2 /L to 4.97 m 2 /L at a constant temperature of 19 ± 2 °C. The concentration of plutonium in solution was 6.0(± 1.0) × 10 −10 M. Fractions of Pu(IV), (V) and (VI) as a function of time were determined by removal of plutonium from the solid phase followed by an ultrafiltration/solvent extraction procedure using TTA and HDEHP as extractants. Appropriate removal conditions were established for Pu(IV), Pu(V) and Pu(VI) using Th(IV), Np(V) and U(VI) as oxidation state analogs. In the pH range from 2.0 to 3.5, the oxidation of Pu(IV) by manganese dioxide was first order with respect to the concentration of manganese dioxide and −0.21 with respect to the hydrogen ion concentration. Consequently, assuming a first order dependence with respect to the concentration of Pu(IV), the oxidation reaction can be described by the following rate equation: with k = 3.72(± 0.13) × 10 −3 (m −2 L)(mol −1 L) −0.21 (min) −1 . Using the kinetic data determined in this study, the influence of manganese dioxide on the oxidation state distribution of plutonium under various environmental conditions as well as in waste solutions can be estimated.

The solubility behavior of uranium dioxide was studied under oxidizing and reducing conditions in simulated Belgian Boom clay water at 25 °C, on unirradiated fuel pellets. For a comparable period (1 month), the uranium concentration and the dissolution rate (10 −12 mol m −2 s −1 ) in reduced medium were ∼30 times smaller than in oxidizing conditions (3 × 10 −11 mol m −2 s −1 ). X-ray photoelectron spectroscopy analyses revealed that U(VI) and U(IV) oxidation states were present in the altered layer in both cases. In the case of the experiments performed in oxidizing conditions, this technique also revealed the formation of U-OH bonds. Although steady state was not reached, geochemical calculations allowed us to suggest a class of secondary phases able to control the uranium solubility in the long term.

A rapid, single-column extraction chromatographic method using commercially available UTEVA resin has been developed for the separation of uranium, plutonium, neptunium and americium. The method yields recoveries superior to 90% and allows direct loading of separated fractions on filaments for subsequent analysis by thermal ionization mass spectrometry. The use of reagents compatible with robotized equipment allows automation of the separation process for routine analysis of nuclear materials. The redox reactions between plutonium, neptunium and hydrogen peroxide involved in the separation process were studied by UV/Vis/NIR absorption spectroscopy.

Extraction equilibrium of lanthanides between tri-butyl phosphate solution and molten calcium nitrate hydrate, which is generally called a hydrate melt, was studied radiochemically. The dependence of the extractions on the water content of the melt was studied at 70 °C in the range of R from 3 to 12.6 for Ca(NO 3 ) 2 · R H 2 O. From the observed distribution ratios, variation of the activity coefficients of lanthanide nitrates in hydrate melts was analyzed as a function of water content. It was found that, in a concentration region of R greater than about 5.7, the increase of the activity coefficients with decreasing R shows a different tendency compared with the more water-abundant region. This is discussed in conjunction with the characteristics of the hydrate melt, as well as with the difference of the ionic size of lanthanide cations.

Single crystals of zircon doped with 239 Pu, up to 2 mm in size, were grown for the first time ever, using a Li-Mo flux. The crystals were transparent and free of inclusions of separated Pu phases. The zircon crystals were studied using optical and scanning electron microscopy, the electron microprobe, and cathodoluminescence imaging and spectroscopy. The incorporation of 239 Pu ranged from 0.1 to 1.4 wt.% el. The intensity of the CL emission in the Pu-doped crystals is correlated with the Pu content.

In the post-closure period of a deep disposal facility for low- to intermediate-level radioactive waste, highly alkaline pore-fluids chemically equilibrated with cementious components of the engineered barriers will migrate into the surroundings. Primary minerals in the host rock could dissolve and secondary calcium silicate hydrate (CSH) phases might be formed in the micro pore system of the rock and change the available porosity for radionuclides and hence the permeability. Since the retardation of radionuclides in the host rock depends on sorption and matrix diffusion, the study of the mineral reactions is of importance. Experiments with diffusion of HTO and synthetic cementitious pore-waters through 1 cm thick discs of Äspö diorite have been made in a nitrogen-flushed glovebox facility. Synthetic cementitious pore-waters representative of fresh and leached concrete were used. For the fresh pore-water the results show that hydroxide ion diffusion is retarded relative to HTO tracer, indicating that reactions between hydroxide and host rock take place. The result can be interpreted as hydroxide sorption in the rock. Al and Si did also accumulate on the sampling side which was attributed to mineral dissolution. For the evolved pore-water, no through-diffusion of hydroxide ions or accumulation of other elements in the sampling cell was observed during the sampling period, indicating that the through-diffusion of hydroxide is hindered by this type of pore-water.

The safe handling and storage of radioactive materials require an understanding of the effects of radiolysis on those materials. Radiolysis may result in the production of gases ( e.g. , corrosives) or pressures that are deleterious to storage containers. A study has been performed to address these concerns as they relate to the radiolysis of residual fluoride compounds in uranium oxides. Samples of UO 2 F 2 · x H 2 O and U 3 O 8 (with ∼1.4 wt.% fluorine content) were irradiated in a 60 Co source and in spent nuclear fuel (SNF) elements from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. Container pressures were monitored throughout the irradiations, and gas and solid samples were analyzed after the irradiations. The irradiation of UO 2 F 2 · x H 2 O produced O 2 – with G ( O 2 )-values ranging from 0.007 to 0.03 molecules of O 2 produced per 100 eV. Neither F 2 nor HF was produced by the irradiations. Chemical analyses of solid samples showed that some of the uranium was reduced from U(VI) to U(IV). A saturation damage limit for the UO 2 F 2 · x H 2 O was demonstrated by using the HFIR SNF elements, and the limit was found to be 7–9% at ∼10 8 rad/h). It is shown that the covalently bonded oxygen is more susceptible to radiation damage than is the ionically bonded fluorine. Irradiation of U 3 O 8 (with ∼1.4 wt.% fluorine content) resulted in neither gas production nor a pressure increase. These experiments led to the conclusion that during long-term storage U 3 O 8 is safe from overpressurization and the production of corrosives caused by gamma radiolysis of residual fluorides.