[en] The fuel elements from nuclear research reactors consist in a laminated sandwich of aluminum with a core of some uranium compound. To process this material its necessary to previously eliminate the aluminum covering the fuel, before the conditioning of the rest of the fuel in a stable matrix, in order to obtain an acceptable waste form for a subsequent disposition in a geological repository. Normally, mechanical and chemical methods are proposed for that purpose. One of the most developed techniques for immobilization of the radioactive elements above mentioned, is the vitrification. In this work we propose a method named CERUS (in Spanish Ceramizacion de Elementos Radiactivos con Uranio Sinterizado - Ceramization of radioactive elements with sintered uranium). This is a sinterization of the pre-treated fuel elements mixed with natural uranium oxide. The properties of the blocks obtained are adequate for final disposal in a deep geological reservoir. (author)

[en] The alpha radiation and the in-pile during irradiation effects on the hardness, the crack nucleation and the fracture toughness of the German alumino borosilicate glass SG7 were investigated by using the Vickers indentation. Cold pressed and sintered samples were irradiated with thermal neutrons, in the Argentine nuclear reactors RA-3 and RA-6, to produce alpha particles in the whole volume of the glass by means of the (n, alpha)-reaction with B-10. The Vickers hardness, the crack nucleation, as 50 percent fracture probability load, plotted as the Weibull's fracture probability distribution function and the fracture toughness, as critical stress intensity factor K_Ic, were correlated to the four cumulative disintegration values. It was ascertained that: a) the Vickers hardness decreases from 5.6 GPa for the non-irradiated sample up to 4.7 GPa for the sample irradiated 70 h at the lower neutron flux (4.0 x 10 - sup 18 - alpha disintegration per cm - sup 3 -), b) the 50 % fracture probability load increases from 1.4 N for the non-irradiated sample up to 4.7 g for the sample irradiated 22 h at the higher flux (6.8 x 10 - sup 18 - alpha disintegration per cm - sup 3 -), and c) the stress intensity factor increases from 0.80 MPa.m - sup 1/2 - for the non irradiated sample up to 0.86 MPa.m - sup 1/2 - for the sample mentioned in b). The in-pile annealing was analyzed by comparing the crack nucleation after irradiation with data obtained by heavy ion irradiation followed by thermal annealing. Results for the SG7 glass were compared to those for soda-lime and borosilicate glasses. (author)

[en] Full Text: The ion exchange resins are used to retain the radionuclides that contaminate the water in primary and secondary circuits and storage pools of the Argentine nuclear reactors. Once used, this resins are an intermediate level waste. Due to the generated volume of resins during the reactor life, it is necessary to have a proper method for management and final disposal of these wastes. Up to now in National Atomic Energy Commission (CNEA), the most studied process is cementation. However, this method increases the waste volume and the final product has low compression hardness. The immobilization in glass of these wastes is attractive because of the volume reduction that could be attained and because of the well known durability of glass. In this work we prepared a mixed bed of resins, similar to those used in Argentina nuclear reactors. We use cesium as a simulant for the active elements present in the resins. Absorption of lithium and cesium was controlled by conductivity and/or ph measurements. The so charged resins were thermally decomposed. This process was studied by Dta/T G experiments. Some potentially problematic effects were founded (foam formation, particle explosion), regarding the possibility of immobilisation of the resins in glass by sintering. Finally, the calcination products were analyzed by Sem and X-ray diffraction. This analysis showed that lithium and cesium remain as sulfates. For this reason we decided to use those chemical compounds as simulant s of the calcination products in the following sintering experiments

[en] The objective of the work was to evaluate the long-term capacity of sintered glass to retain the high-level nuclear wastes (HLW) in repository conditions. The SG7 glass containing simulated non-radioactive waste PHWR was irradiated in a Co⁶⁰ γ-field (Division Fuentes Intensas - Centro Atomico Ezeiza) at dose rate of 43.4K Gy/h. We were considered the following potential effects of irradiation: changes in the microhardness and in the density; changes in leaching and dissolution rates at 363 k during 3 days and 28 days, in deionized; changes in the microstructure of the sintered glass forms. Globally we did not find any adverse influence on the bulk corrosion resistance or in the microstructure. We were calculated, by means of the computer code ORIGEN2.1, the fission products activities caused by the high-level wastes from reprocessed spent fuel type PHWR, with a burnup of 7000 MWd/t U and a cooling time of 20 years. We presented the predicted numbers of cumulative gamma events, the activity and the dose due to gamma radiation, per metric to of heavy metal and per cubic centimeter of the sintered HLW glass forms, as a function of the time up to one million years. (author)

[en] We have subjected the borosilicate glass to thermal neutron irradiation in a reactor, with an accumulated fluence equivalent to approximately E3, E4, E5, y E6 years of waste disposal. We considered the following potential effects of accumulated alpha decay: a) Changes in the density; b) Changes in the dissolution rates; c) Changes in the microstructure of the sintered glass. (author)

[en] Borosilicate german glass SG7 samples, obtained by frit sintering, were irradiated with different fluences of thermal neutrons in the nucleus of a nuclear reactor. The nuclear reaction ¹⁰ B(n,α)⁷ Li, where the ¹⁰ B isotope is one of the natural glass components, was used to generate alpha particles throughout the glass volume. The maximum alpha disintegration per unit volume achieved was equivalent to that accumulated in a borosilicate glass with nuclear wastes after 3.8 million years. Through Vickers indentations values for microhardness, stress for 50% fracture probability (Weibull statistics) and estimation of the toughness were obtained as a function of alpha radiation dose. Two counterbalanced effects were found: that due to the disorder created by the alpha particles in the glass and that due to the annealing during irradiation (temperature below 240 deg C). Considering the alpha radiation effect, glasses tend decrease Vickers hardness, and to increase thr 50% fracture probability stress with the dose increase. (author)

[en] Muon tomography is a technique based on the observation of the absorption of muons in matter, as the ordinary radiography does using X-rays. The interaction of cosmic rays with the atmosphere produce Extensive Air Showers, which provide an abundant source of muons. These particles can be used for multiple applications on muon tomography, in particular, to study the internal structure of different volcanoes edifices. It is necessary in any muon tomography experiment a particle detector that has the capability to be sensitive to muon interactions. Here we report a brief description of the first steps towards a complete design of a prototype particle detector to perform muon tomography in harsh conditions encountered in the surroundings of the Galeras Volcano. The mechanical design and fabrication processes of the supporting structure of the muon detector prototype, and first steps towards a future detector simulations on GEANT4 are described in this work. (paper)